ML20363A116

2018 Donald C. Cook Nuclear Power Plant Exam Proposed Written

ML20363A116
Person / Time
Site:	Cook
Issue date:	08/13/2018
From:	Chuck Zoia NRC/RGN-III/DRS/OLB
To:	Indiana Michigan Power Co
Zoia C
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ML17164A386	List: ML20363A070 ML20363A094 ML20363A097 ML20363A098 ML20363A104 ML20363A114 ML20363A116 ML20363A131
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Download: ML20363A116 (201)

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 1 ID: CM-40268 Points: 1.00 Unit 2 was operating at steady state full power when a loss of off-site power occurred. The following indications were observed during the performance of Step 1 of 2-OHP-4023-E-0, Reactor Trip or Safety Injection:

• WR Neutron flux is 3% and lowering

• Prior to RCP Bus transfer, the operator noted that Rod H8 was at 50 steps

• RTB is closed

• RTA, BYA, and BYB are open

• All Auxiliary Feedwater Pumps are Running The above indications remained constant when the operators actuated the manual reactor trip breaker switch.

Which one of the following actions should the crew take?

A. Go to 2-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS B. Go to 2-OHP-4023-ECA-0.0, Loss of all AC Power C. Continue in 2-OHP-4023-E-0, Reactor Trip or Safety Injection D. Go to 2-OHP-4023-FR-S.2, Response to Loss of Core Shutdown Answer: C Answer Explanation:

A. INCORRECT. Plausible because RTB B remains closed and rod H-8 isstuck at 50 steps.

Incorrect as FR-S.1 is used if WR Flux is not less than 5% or if flux is rising.

B. INCORRECT. Plausible because there is a loss of off-site power. Incorrect as ECA-0.0 is entered from a loss of ALL AC power, not just a loss of offsite power.

C. CORRECT. All trip breakers are not open, the operator continues with E-0 Step 1 RNO, verifies the trip by WR flux less than 5% and lowering, and then continues with E-0 Step 2.

D. INCORRECT. Plausible because rod H-8 isstuck at 50 steps. Incorrect as FR-S.2 is optionally entered (Yellow Path) on WR Startup rate not being as low as expected for conditions. These conditions do not meet the entry conditions for FR-S.2.

DC COOK OPS Page: 1 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: CM-40268 External Topic ID: PROC-3: Procedure Knowledge Comments:

REFERENCE:

2-OHP-4023-E-0 Reactor Trip or Safety Injection K/A 000007 EA2.02 Reactor Trip - Stabilization Ability to determine and interpret the following as they apply to a reactor trip:

Proper actions to be taken if the automatic safety functions have not taken place.

RO - 4.3 SRO - 4.6 CFR - 41.7 / 41.10 / 43.5 / 45.13 K/A Match Candidate must determine if the plant is responding correctly even though a trip breaker has failed to open.

Associated objective(s):

(RO-C-EOP03-E14) List the Immediate Operator Actions for E-0.

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 2 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 2 ID: RQ-C-4061-T1-01 Points: 1.00 Unit 1 is at 100% Power. The Pressurizer Relief Tank Pressure High or Low annunciator just came in. The Crew is looking to enter either 1-OHP-4022-002-009, Leaking Pressurizer Power Operated Relief Valve or 1-OHP-4030-102-016, Reactor Coolant System Leak Test.

Which indication(s) would be the most helpful in identifying which procedure to enter for this condition?

1) Common Pressurizer PORV Tailpipe Temperature

2) Individual Pressurizer PORV Tailpipe Temperature

3) Common Pressurizer PORV Acoustic Monitor

4) Individual Pressurizer PORV Acoustic Monitor

5) Common Pressurizer Safety Tailpipe Temperature

6) Individual Pressurizer Safety Tailpipe Temperature

7) Common Pressurizer Safety Acoustic Monitor

8) Individual Pressurizer Safety Acoustic Monitor A. 1,4,5,8 B. 1,3,6,8 C. 2,4,5,7 D. 2,3,6,7 Answer: B Answer Explanation:

A. INCORRECT - The Pressurizer PORVs all share a common tailpipe Temperature probe and a common Acoustic Monitor. Whereas the Pressurizer Safety Valves all have individual Temperature probes and individual Acoustic Monitors.

B. CORRECT - The Pressurizer PORVs all share a common tailpipe Temperature probe and a common Acoustic Monitor. Whereas the Pressurizer Safety Valves all have individual Temperature probes and individual Acoustic Monitors.

C. INCORRECT - The Pressurizer PORVs all share a common tailpipe Temperature probe and a common Acoustic Monitor. Whereas the Pressurizer Safety Valves all have individual Temperature probes and individual Acoustic Monitors.

D. INCORRECT - The Pressurizer PORVs all share a common tailpipe Temperature probe and a common Acoustic Monitor. Whereas the Pressurizer Safety Valves all have individual Temperature probes and individual Acoustic Monitors.

DC COOK OPS Page: 3 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RQ-C-4061-T1-01 Comments:

REFERENCE:

SOD 00202-001 Over Pressure Protection Systems Significantly modified CM-7061 K/A 000008 2.4.3 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open)

Ability to identify Post-Accident Instrumentation RO - 3.7 SRO - 3.9 CFR - 41.6 / 45.4 K/A Match Candidate must demonstrate knowledge of instrument indications available to monitor the plant in a Post - Accident condition.

Associated objective(s):

(RQ-C-4061-T1) Operate the Pressurizer and Pressure Relief System in accordance with applicable plant procedures.

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 4 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 3 ID: NRCAUDIT07-1068 Points: 1.00 A Large Break LOCA has occurred. The following conditions exist:

• RCS pressure is 20 psig and slowly lowering

• RCS Cold Leg and CETC temperatures are 280oF

• Aux. Feedwater flow is 450 x 103 pph

• WR Level in all FOUR (4) SGs indicates 55%

• SG Pressures are 800 psig What is the source of core heat removal?

A. Core reflux cooling B. Steam relief out the SGs C. Break flow D. Natural circulation flow Answer: C Answer Explanation:

A. INCORRECT - Plausible if student believes that reflux cooling continues in a LBLOCA.

Incorrect as reflux cooling terminates as the RCS temperature lowers to < SG saturation temperature (~515oF).

B. INCORRECT - Plausible if student believes that SGs continue to remove heat throughout a LBLOCA. Incorrect as RCS temperature is lower than SG saturation temperature (~515oF).

C. CORRECT - The major method of heat transfer from the fuel occurs as the RCS flashes to steam. For LBLOCAs the SGs become ineffective as a heat removal method early on in the accident. As RCS pressure continues to drop, the SGs begin to act as a heat source. The Accumulators begin to inject. During the blowdown phase all ECCS and Accumulator flow is assumed to go out the cold leg break and bypass the core entirely. The Blowdown phase ends when RCS pressure falls to a value approaching that of containment pressure.

Two-phase blowdown takes place for the majority of the time that RCS pressure is lowering. Both core uncovery and termination of natural circulation occur because of the net loss of RCS water inventory through the break. Once the SG tubes become completely steam-bound, natural circulation is terminated.

D. INCORRECT - Plausible if student believes that two phase natural circulation occurs throughout LBLOCA. Incorrect as once the SG tubes become completely steam-bound, natural circulation is terminated. RCS temperature is lower than SG saturation temperature (~515oF).

DC COOK OPS Page: 5 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-1068 Comments:

REFERENCE:

RO-C-EOP09 pg. 34-41 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 K/A Match 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.

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 6 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 4 ID: RO-C-00201-E6-1 Points: 1.00 Unit 1 is in Mode 3 at normal operating temperature and pressure.

While performing trouble shooting activities, CCM-453 (CCW from RCP Thermal Barrier) closes.

Which ONE of the following describes the impact this failure will have on the Reactor Coolant Pump(s) (RCP)?

Thermal Barrier Component Cooling Water return flow has been lost from:

A. only 1 RCP. RCP operation may continue.

B. all RCPs. RCP operation may continue.

C. only 1 RCP. The affected RCP must be stopped within 2 minutes due to a loss of cooling.

D. all RCPs. All RCPs must be stopped within 2 minutes due to a loss of cooling.

Answer: B Answer Explanation:

A. INCORRECT - The thermal barrier isolation valve is common to all RCPs. Plausible since NESW cooling has separate valves.

B. CORRECT - The CCW thermal barrier isolation valve is common to all RCPs. RCP operation may continue provided seal injection is maintained.

C. INCORRECT - The CCW thermal barrier isolation valve is common to all RCPs. Plausible since NESW cooling has separate valves and a loss of CCW BEARING cooling does require the RCPs to be stopped.

D. INCORRECT - The CCW thermal barrier isolation valve is common to all RCPs. Plausible since NESW cooling has separate valves and a loss of CCW BEARING cooling does require the RCPs to be stopped.

Question ID: RO-C-00201-E6-1 External Topic ID: 0201: Reactor Coolant Pump Level/Difficulty: H/2 Comments:

REFERENCE:

SOD-01600-002, K/A 000015 AK2.08 Reactor Coolant Pump (RCP) Malfunctions Know of interrelations between the Reactor Coolant Pump Malfunctions and the following:

CCWS RO - 2.6 SRO - 2.6 CFR - 41.7 / 45.7 DC COOK OPS Page: 7 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A Match Candidate must demonstrate knowledge of the CCWS and RCP interrelation when evaluating a Loss of CCW flow to a RCP and whether continued operation is allowed IAW RCP Malfunction procedure.

Associated objective(s):

(RO-C-00201-E6) Explain how a loss of each support system will affect RCP operation:

a. CVCS

b. CCW

c. NESW/Chill Water

d. Total loss of seal cooling (CVCS seal injection along with CCW to Thermal Barrier)

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 8 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 5 ID: 0050230801-E1-5 Points: 1.00 Given the following Unit 1 conditions:

• The Unit is at 100% power.

• Annunciator 108 Drop 4 (PZR LO LEVEL DEVIATION) is received.

• Pressurizer level is currently 41% and slowly decreasing.

• 1-QRV-251 valve demand signal indicates 100%.

• 1-QRV-251 is in auto.

Legend:

• 1-QRV-200 - Charging to Regenerative Heat Exchanger Flow Control Valve

• 1-QRV-251 - CVCS Charging Pump Discharge Flow Control Valve Assuming NO operator action which ONE of the following describes the cause of this transient and what is the status of RCP Seal Injection flow?

A. QRV-200 has failed Open Seal injection has max flow B. QRV-200 has failed Closed Seal injection has max flow C. QRV-200 has failed Open Seal injection has min flow D. QRV-200 has failed Closed Seal injection has min flow Answer: B Answer Explanation:

A. INCORRECT - QRV-200 is normally throttled to control Seal Injection flow by supplying back pressure to the normal charging flow path. Plausible if candidate confuses valve function with QRV-251. Second part is indicative of QRV-200 being full closed not open.

B. CORRECT - QRV-200 will isolate the normal charging line flowpath and direct all flow thru the RCP seal line.

C. INCORRECT - QRV-200 failing full open would result in full charging flow controlled by QRV-251 and minimum Seal Injection. This would result in a high Pressurizer level.

D. INCORRECT - The first part is true but the Seal Injection would have maximum flow for the second part.

DC COOK OPS Page: 9 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: 0050230801-E1-5 Level/Difficulty: H/3 Comments:

REFERENCE:

1-OHP-4024-108 Drop 4, SOD-00300-001 Charging and Letdown K/A 000022 AK3.01 Loss of Reactor Coolant Makeup Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup:

Adjustment of RCP seal backpressure regulator valve to obtain normal flow.

RO - 2.7 SRO - 3.1 CFR - 41.5 / 41.10 / 45.6 / 45.13 K/A Match:

Given indications of reduced Rx Coolant makeup, the applicant is required to determine the cause of the failure and interpret the associated effect on RCP seal injection supply flow.

Associated objective(s):

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 10 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 6 ID: 2008NRC-0639 Points: 1.00 Unit 2 has experienced a loss of both CCW pumps in MODE 3.

Given the following plant conditions:

• NEITHER Unit 2 CCW pump can be restarted

• Unit 2 East CCP is running

• Unit 2 West CCP is in standby

• 2-OHP-4022-016-004, Loss of Component Cooling Water, is in progress Which ONE of the following describes the procedural requirements for CCP operation based on these conditions?

A. Immediately stop the East CCP until the Unit 1 CCW system has been cross-tied.

B. Run the East CCP for up to 5 minutes and then stop the East CCP and initiate a CVCS cross-tie to Unit 1.

C. Run the East CCP as long as it continues to operate or until a CVCS cross-tie to Unit 1 can be performed.

D. Run the East CCP as long as it continues to operate and then start the West CCP at minimum flow from the RWST until the Unit 1 CCW system has been cross-tied.

Answer: C Answer Explanation:

A. INCORRECT - One pump should be run as long as possible to allow time to align CVCS crosstie.

B. INCORRECT - One pump should be run as long as possible to allow time to align CVCS crosstie. (The pump may trip after 1.5 minutes)

C. CORRECT - 2-OHP-4022-016-004 has a note prior to step 4 that describes the possible damage that may occur to a CCP on the loss of CCW. The note and procedure directs that one CCP be saved until CCW is restored. The other pump should be run as long as possible to allow time to align CVCS crosstie D. INCORRECT - One pump should be run as long as possible to allow time to align CVCS crosstie. (The pump may trip after 1.5 minutes) The running pump is aligned to the RWST with minimum flow and then Unit 1 crosstie for CCW is established.

DC COOK OPS Page: 11 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: 2008NRC-0639 External Topic ID: PROC-3: Procedure Knowledge Comments:

REFERENCE:

2-OHP-4022-016-004 Loss of Component Cooling Water K/A 000026 AA1.02 Loss of Component Cooling Water (CCW)

Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water:

Loads on the CCWS in the control room RO - 3.2 SRO - 3.3 CFR - 41.7 / 45.5 / 45.6 K/A Match Candidate must demonstrate ability to monitor loads cooled by CCW and also demonstrate when operation is permitted and prohibited following a Loss of CCW.

Associated objective(s):

(RO-C-AOP0560412-E2) explain the required operator actions to stabilize plant conditions after a Loss of CCW prior to formal procedure implementation (RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 12 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 7 ID: RO-C-AOP0330412-E3-1 Points: 1.00 Given the following conditions:

• Unit 2 is operating at 70% power.

• Panel 208, Drop 7; PZR PRESS HIGH DEVIATION is received in the control room.

• Pressurizer Pressure Transmitter NPP-151 indicates 2310 psig and RISING.

• Pressurizer Pressure Transmitter NPP-152 indicates 2225 psig and LOWERING.

Which one of the following describes the failure and the actions required to correct this failure?

A. NPP-152 is failing low, place Pressurizer Master Pressure Controller in manual, raise demand to restore pressure, and select Channel 4 for Control.

B. NPP-151 is failing high, place Pressurizer Master Pressure Controller in manual, lower demand to restore pressure, and select Channel 3 for Control.

C. The Pressurizer Bistable channel has failed, de-energize the pressurizer heaters to restore pressure and select Channel 4 for the Bistable channel.

D. The Pressurizer Master Pressure Controller is failing high, place Pressurizer Sprays in manual, raise demand to restore pressure, and control the Sprays and Heaters in Manual.

Answer: B Answer Explanation:

A. INCORRECT - NPP-152 is the bistable channel. If it failed low the pressure would not rise.

Channel 4 is the backup for channel 2 B. Correct - NPP151 failing high will cause the demand on the master controller to rise opening the sprays and lowering actual pressure. Channel 3 is the backup for the control channel.

C. INCORRECT - The bistable channel (NPP-152) failing low will not impact the Heaters.

(Plausible since the Level channels both impact the Heaters) Channel 4 is the backup for channel 2 Bistable channel.

D. INCORRECT - If the master controller failed high the response would be for the alarm, and sprays to open. The actions taken would be correct for this failure. The NPP-151 indication would also be lowering.

DC COOK OPS Page: 13 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-AOP0330412-E3-1 Level/Difficulty: H/3 Comments:

REFERENCE:

2-OHP-4022-013-009, Pressurizer Pressure Instrument Malfunction, SOD-00202-002, SOD-00202-001 K/A 000027 AA2.18 Pressurizer Pressure Control (PZR PCS) Malfunction Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions:

Operable control channel RO - 3.4 SRO - 3.5 CFR - 41.7 / 41.10 / 43.5 / 45.13 K/A Match Candidate must evaluate plant conditions and use knowledge of controlling channel functions for Pressurizer Pressure to determine actions for a channel malfunction.

Associated objective(s):

(RO-C-AOP0330412-E3) explain the procedural mitigation strategy for a Pressurizer Pressure Instrument Malfunction or Controller Failure (RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 14 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 8 ID: RO-C-EOP08-E18-2 Points: 1.00 Given the following plant conditions on Unit 2:

• A tube rupture has occurred in the SG #23.

• A Manual Reactor Trip and Safety Injection were performed.

• RCS Pressure is currently 2230 psig.

• RCS Tavg is 549°F.

• SG#23 has been isolated in accordance with procedure 2-OHP-4023-E-3, Steam Generator Tube Rupture.

• SG#23 Pressure is 995 psig.

The Unit Supervisor has directed you to perform the cooldown to 488°F using the Steam Dumps.

Which ONE of the following describes how the cooldown would be performed and the reason?

A. Open the Steam dumps to obtain a cooldown rate of < 200°F/Hr and depressurize the RCS to maintain <1200 psid SG Tube Differential Pressure. Complete the cooldown to allow a primary depressurization to limit the SG tube DP.

B. Open the Steam dumps to obtain the Maximum Rate possible since ruptured SG Pressure is >450 psig. Complete the cooldown to allow a primary depressurization to limit the SG tube leakage.

C. Open the Steam dumps to obtain a cooldown rate of < 100°F/Hr to prevent exceeding Technical Specification limits.

D. Open the Steam dumps to obtain a cooldown rate of < 60°F/Hr to prevent a Steamline Isolation.

Answer: B Answer Explanation:

A. INCORRECT - Steam Dumps are operated to obtain the maximum rate possible. 200°F/Hr may be the maximum achievable rate but the SG DP limit is 1600 PSID and depressurization is not performed during Cooldown.

B. CORRECT - The dumps should be opened in a controlled manner to obtain the maximum cooldown rate. SG Pressure has been previously checked >450 psig to prevent a PTS concern. The TS limit of 100 °F/Hr may be exceeded.

C. INCORRECT - Obtain the maximum cooldown rate. SG Pressure has been previously checked >450 psig to prevent a PTS concern. The TS limit of 100 °F/Hr may be exceeded.

DC COOK OPS Page: 15 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. INCORRECT - The dumps should be opened in a controlled manner to obtain the maximum cooldown rate. Steam Line Isolation is blocked as temperature lowers.

Question ID: RO-C-EOP08-E18-2 Level/Difficulty: H/3 Comments:

REFERENCE:

02-OHP-4023-E-3 Step 6 & 7 and Background, RO-C-EOP08 K/A 000038 2.2.22 Steam Generator tube Rupture Knowledge of limiting conditions for operational and safety limits RO - 4.0 SRO - 4.7 CFR - 41.5 / 43.2 / 45.2 K/A Match Candidate must demonstrate knowledge of EOP Cooldown rate required which may exceed the TS limits.

Associated objective(s):

(RO-C-EOP08-E18) For the E-3 series procedures and the ECA-3 series procedures discuss the basis or reason for all Steps.

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 16 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 9 ID: RO-C-EOP07-E8-10 Points: 1.00 The Unit Supervisor has just entered 02-OHP-4023-ECA-2.1, Uncontrolled Depressurization of All Steam Generators.

• The Turbine Driven Aux Feedwater Pump is the only AFW pump running.

• No AFW flow adjustments have been made since AFW auto started after the trip.

• AFW Flow is currently 90,000 PPH per SG.

• RCS cold leg temperatures have lowered by 125oF and are still lowering.

• ANN Panel 114 Drop 9 TDAFP Discharge Flow High Alarm is LIT.

• SG Water Level LOW-LOW alarms are Lit for all 4 SGs.

Which ONE of the following actions is required to be taken with the AFW system?

The AFW Flow Retention:

A. does NOT need to be reset to throttle closed the AFW valves to the SGs. Lower AFW flow to 25,000 PPH per steam generator.

B. has cleared and does NOT need to be reset to open the AFW valves to the SGs. Raise AFW flow to 110,000 PPH per steam generator.

C. must be reset to throttle closed the AFW valves to the SGs. Lower AFW flow to 25,000 PPH per steam generator.

D. must be reset to throttle open the AFW valves to the SGs. Raise AFW flow to 110,000 PPH per steam generator.

Answer: A Answer Explanation:

A. CORRECT - the flow retention does not need to be reset to close the valves. A Minimum flow is maintained to each SG to minimize thermal shock. ECA-2.1 Step 2 directs the Crew to reduce FW Flow to 25 x103 PPH to each SG if the cooldown rate is > 100oF/hr.

B. INCORRECT - The high flow alarm is not cleared but the candidate may think the flow retention is cleared based on current flow rates. The 110 x103 PPH to each SG is the max 450 x103 PPH normally referenced in EOPs when SG levels are low.

C. INCORRECT - the flow retention does not need to be reset to close the valves. A Minimum flow is maintained to each SG to minimize thermal shock. ECA-2.1 Step 2 directs the Crew to reduce FW Flow to 25 x103 PPH to each SG if the cooldown rate is > 100oF/hr.

D. INCORRECT - The 110 x103 PPH to each SG is the max 450x103 PPH normally referenced in EOPs when SG levels are low and a large cooldown is not in progress.

DC COOK OPS Page: 17 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-EOP07-E8-10 Level/Difficulty: H/3 Comments:

REFERENCE:

02-OHP-4023-ECA-2.1 Uncontrolled Depressurization of All Steam Generators, 2-OHP-4024-213 Drop 19 K/A 00WE12 EK1.3 Uncontrolled Depressurization of all Steam Generators Knowledge of the operational implications of the following concepts as they apply to the Uncontrolled Depressurization of all Steam Generators:

Annunciators and conditions indicating signals, and remedial actions associated with the Uncontrolled Depressurization of all Steam Generators RO - 3.4 SRO - 3.7 CFR - 41.8 / 41.10 / 45.3 K/A Match Candidate must demonstrate an understanding of annunciator meaning and actions required due to standing annunciator.

Associated objective(s):

(RO-C-EOP07-E8) Explain AFW flow control during an uncontrolled depressurization of all SGs to include:

a. The four (4) reasons for controlling AFW flow

b. Minimum AFW flow rate to each SG and the basis for this minimum limit (RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 18 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 10 ID: RO-C-01100-E2-2 Points: 1.00 Unit 1 was operating at 85% power when the East MFP Trip followed shortly thereafter by the West MFP Trip.

Which one of the following describes the plant response and reason?

The Loss of Two FW pumps will cause:

A. a direct Reactor trip to provide protection from Loss of heat sink.

B. a Reactor trip from SG Low-Low water levels to limit the RCS pressure/temperature transient.

C. a direct Turbine trip which causes a Reactor trip to limit the RCS pressure/temperature transient.

D. a Reactor trip due to Low SG level coincident with Low FW Flow to provide protection from Loss of heat sink.

Answer: D Answer Explanation:

A. INCORRECT - The reactor is not directly tripped from loss of 2 FW pumps.

B. INCORRECT - The reactor would trip from Low 2 SG levels if it did not trip from the SG low with low FW flow first. The reason given is based on the loss of heat removal/turbine trip.

C. INCORRECT - Loss of 2 FW pumps will start the AFW pumps but not directly trip the turbine.

This is the reason for the turbine trip/reactor trip.

D. CORRECT - The reactor trips on a low level with low FW flow to provide loss of heat sink protection.

Question ID: RO-C-01100-E2-2 Level/Difficulty: F/2 Comments:

REFERENCE:

RQ-C-Know DC COOK OPS Page: 19 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 000054K3.01 Loss of Main Feedwater (MFW)

Knowledge of the reasons for the following responses as they apply to the Loss of Main Feedwater (MFW):

Reactor and/or turbine trip, manual and automatic K/A Match Candidate must understand the plant response to a Loss of Main Feedwater and determine if any reactor or turbine trips are associated with the loss.

Associated objective(s):

(RO-C-01100-E2) Explain the purposes/functions/bases of the following RPS actuation circuits (Reactor Trips):

a. Manual

b. Power Range (high, low, and rate)

c. Intermediate Range

d. Source Range

e. OT(T

f. OP(T

g. Pressurizer Pressure (high and low)

h. Pressurizer Level

i. Loss of RCS Flow

j. SG Low-Low Level

k. SG Low Level w/ Feed/Steam Flow Mismatch

l. RCP UV

m. RCP UF

n. RCP Breaker Position

o. Turbine Trip (oil pressure and stop valve position)

p. Safety Injection (RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 20 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 11 ID: CM-7897 Points: 1.00 Which ONE of the following interlocks must be satisfied to close in the Emergency Power feed breaker to Bus T11A?

A. Bus T11A tie breaker from bus 1A must be open and Bus T11B tie breaker from bus 1B must be open.

B. Bus T11A tie breaker from bus 1A must be open and AB diesel to T11A output breaker must be open / Lockout.

C. TR12EP-1 lockout relay must be reset and the DG 1AB lockout (HEA) relay must be actuated.

D. EP Motorized Disconnect Switch must be open and both the supplemental diesels running supplying EP.

Answer: B Answer Explanation:

A. INCORRECT - RCP feed from Bus 1A must be open, but the position of RCP feed from Bus 1B to T11B is not interlocked. Plausible if student believes that the RCP tie breakers for each bus on the train must be open in order to close in EP.

B. CORRECT - Both other feed breakers must be open.

C. INCORRECT - The EP system can be supplied from the SDGs even with the transformer locked out; transformer lockout is not an interlock; the DG HEA is not an interlock. Plausible if the student believes the DG feed interlock is provided by the DG HEA, rather than the open status of the DG output breaker for the bus.

D. INCORRECT - The EP breaker can be closed whether EP is fed from the grid or from the SDGs; there is no interlock from SDG status. Plausible if student infers that the EP feed breaker can only be closed if SDGs are supplying the bus.

Question ID: CM-7897 External Topic ID: 8200: Electrical Distribution Level/Difficulty: F / 3 Comments:

REFERENCE:

RO-C-08200 BOP Electrical DC COOK OPS Page: 21 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 000055 EA2.06 Loss of Offsite and onsite power (Station Blackout)

Ability to determine and interpret the following as they apply to a Station Blackout:

Faults and lockouts that must be cleared prior to re-energizing buses.

RO - 3.7 SRO - 4.1 CFR - 41.7 / 41.10 / 43.5 / 45.13 K/A Match Candidate must know the interlocks (lockouts) that must be cleared before the feed breaker can be closed.

Associated objective(s):

(RO-C-08200-E3) Describe interlocks and/or closing conditions which must be satisfied for breaker closure:

a. Reserve Feed Breakers

b. Generator output breakers, including closure for backfeed

c. RCP bus fast transfer closure of reserve feed breakers

d. Supplemental Diesel Generator Output Breakers and Tie Breaker

e. Emergency power breakers on 4kV safety buses (T buses)

f. 600V BOP Cross-tie breaker

g. RCP bus normal and reserve feed breakers (RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 22 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 12 ID: CM-0217A 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 T21D.

Which ONE of the following describes a correct response of the U2 East and U2 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, and then the Blackout Sequencer will start the East CCW pump.

The West CCW pump will not start.

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

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: C Answer Explanation:

A. INCORRECT - The East Pump supplied from T21D and is shed from the bus. The West CCW pump auto starts from low pressure.

B. INCORRECT - The East Pump supplied from T21D and is shed from the bus and restarts on blackout sequencer. The west will start on Low CCW Header Pressure.

C. CORRECT - The East Pump supplied from T21D and is shed from the bus and restarts on blackout sequencer. The west will start on Low CCW Header Pressure.

D. INCORRECT - The East Pump supplied from T21D and is shed from the bus and restarts on blackout sequencer. The West CCW pump auto starts from low pressure.

DC COOK OPS Page: 23 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: CM-0217A External Topic ID: 1600: Component Cooling Water Level/Difficulty: H/3 Comments:

REFERENCE:

4024-104 Drop 99 1-OHP-4021-132-217A DG1CD Load Sequencing &

ESF Testing K/A 000056 AA1.09 Loss of Offsite power Ability to operate and / or monitor the following as they to the Loss of Offsite Power:

CCW pump RO - 3.3 SRO - 3.3 CFR - 41.7 / 45.5 / 45.6 K/a Match 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.

Original Question # - RO28 AUDIT, CM-0217 Original Question KA - 008 A3.08 Modified question to change to bus T21D also changed answers B & C to ensure the response is correct and complete.

Changed correct answer from D to C Associated objective(s):

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 24 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 13 ID: RO-C-03200-E10-03 Points: 1.00 The AB Emergency Diesel Generator was started and paralleled to its respective bus.

Following a loss of AB 250 VDC power, which ONE of the following describes the status of the Emergency Diesel Generator?

A. The DG will continue to run with only the Mechanical Overspeed Trip available.

B. The diesel will trip immediately due to the loss of DC power to the fuel solenoids.

C. The diesel will trip immediately on mechanical overspeed due to loss of DC power to the governor.

D. The DG will continue to run with ONLY the Emergency trips still functional.

Answer: A Answer Explanation:

A. CORRECT - The DG will continue to run but only the mechanical overspeed trip still works.

B. INCORRECT - Plausible if student believes that trips are de-energize to actuate especially since the DG was not in emergency mode.

C. INCORRECT - While the Mechanical overspeed is the only trip available the DG will not overspeed on the loss of DC to the governor.

D. INCORRECT - Plausible if student believes that emergency trips do not require DC power.

Question ID: RO-C-03200-E10-03 External Topic ID: 3200: Diesel Generators Comments:

REFERENCE:

4024-119 Drop 57, DG1AB Trip Relay Bus Volt Failure, RO-C-03200 EDGs K/A 000058 AK3.01 Loss of DC Power Knowledge of the reasons for the following responses as they apply to the Loss of DC Power; Use of DC Control power by the EGDs RO - 3.4 SRO - 3.7 CFR - 41.5 / 41.10 / 45.6 / 45.13 K/A Match Candidate must understand how DC power is used by the EDGs Control system.

DC COOK OPS Page: 25 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Associated objective(s):

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 26 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 14 ID: NRCAUDIT07-0879 Points: 1.00 Given the following conditions:

• Unit 1 is at 100% power.

• The crew has entered 01-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 of the following 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.

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.

DC COOK OPS Page: 27 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0879 Comments:

REFERENCE:

SOD-01900-001 K/A 000062 AA1.06 Loss of Nuclear Service Water Ability to operate and/or monitor the following as they apply to the Loss of Nuclear Service Water (SWS):

Control of flow rates to the components cooled by the SWS.

RO - 2.9 SRO Value 2.9 CFR - 41.7 / 45.5 / 45.6 K/a Match Candidate must be able to monitor which components have lost flow following a loss of ESW Associated objective(s):

(RO-C-01900-E2) Given the Essential Service Water System, describe the flowpath for the system, passing through the following components:

a. ESW pump suction vault

b. ESW pump

c. ESW pump discharge valves

d. ESW pump discharge strainers

e. ESW Unit crosstie valves

f. 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

g. Radiation monitors

h. ESW Return Header isolation valves (RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 28 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 15 ID: NRCAUDIT07-0165 Points: 1.00 During performance of 1-OHP-4022-064-002, Loss of Control Air Recovery, you are preparing to initiate a cooldown and depressurization. The procedure contains a caution pertaining to Safety Injection (SI).

The SI referred to in this caution may be caused by:

A. high containment pressure due to a rupture of the PRT from operation of the PRZ PORVs.

B. steam line differential pressure due to uneven cooling of the Steam Generators.

C. high containment pressure due to a loss of Containment ventilation.

D. low RCS pressure due to loss of Pressurizer pressure control.

Answer: B Answer Explanation:

A. INCORRECT - Use of the PORVs should be limited so PRT pressure will be controllable.

B. CORRECT - If Steam line pressure in any SG becomes >100 psig less than the others a SI signal will be generated.

C. INCORRECT - Loss of air should not significantly impact containment ventilation.

D. INCORRECT - RCS pressure can be controlled through use of backup heaters and PORVs.

Question ID: NRCAUDIT07-0165 External Topic ID: 6401: 6401, Compressed Air Comments:

REFERENCE:

1-OHP-4022-064-002 Loss of Control Air Recovery K/A 000065 2.1.23 Loss of instrument air 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 K/A Match:

Candidate must understand the importance of the procedure steps performed during the Loss of air procedure.

Cognitive Level: H 3 DC COOK OPS Page: 29 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Associated objective(s):

(RO-C-AOP0490412-E3) explain the procedural mitigation strategy for Control Air System Malfunctions/Loss of Control Air Recovery (RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 30 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 16 ID: 2008NRC-0587 Points: 1.00 Given the following plant conditions on Unit 2:

• The unit was operating at 12% power.

• Reserve Aux Transformer 201AB was supplying Buses 2A and 2B.

• Reserve Aux Transformer 201CD was supplying Buses 2C and 2D.

An electrical transient occurred on the Grid.

During the transient the following values were reached:

Electrical Bus 2A 2B 2C 2D Frequency (Hz) 57 57 59.7 59.7 Local Voltage (Volts AC) 4100 4100 4200 4200 CR Indication (Volts AC) 118.2 118.2 121.1 121.1 Ann. 207, Drop 10, RCP BUSSES UNDER FREQ TRIP is LIT.

Which ONE of the following describes the expected status of the reactor trip breakers and the Reactor Coolant Pumps?

Note: Assume no operator action.

A. Reactor Trip Breakers remain Closed All 4 RCPs remain running B. Reactor Trip Breakers Trip Open All 4 RCPs remain running C. Reactor Trip Breakers Trip Open All 4 RCPs trip off D. Reactor Trip Breakers remain Closed ONLY 2 RCPs remain running Answer: B Answer Explanation:

A. INCORRECT - A Reactor trip is generated above 10% (P-7) power when 2/4 buses<58.4 Hz.

(operator may confuse P-7 with P-8 blocks, P-8 associated with 2/4 RCP Low flow).

B. CORRECT - The RCPs no Longer trip on Under frequency (Recent Plant Modification)

C. INCORRECT - A Reactor trip is generated above 10% (P-7) power when 2/4 buses<58.4 Hz.

DC COOK OPS Page: 31 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. INCORRECT - A Reactor trip is generated above 10% (P-7) power when 2/4 buses<58.4 Hz.

(operator may confuse P-7 with P-8 blocks, P-8 associated with 2/4 RCP Low flow).

Question ID: 2008NRC-0587 External Topic ID: 1100-1: RPS/ESFAS Signals Comments:

REFERENCE:

02-OHP-4024-207 Drop 10 RCP Busses Under Freq Trip, RO-C-01100 RPS/ESFAS Signals K/A 000077 AK2.03 Generator Voltage and Electric Grid Disturbances Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following:

Sensors, detectors, indicators RO - 3.0 SRO - 3.1 CFR - 41.4 / 41.5 / 41.7 / 41.10 / 45.8 K/A Match - Question tests knowledge of the sensors used and the coincidence / interlocks for the grid disturbance reactor trip.

Associated objective(s):

(RO-C-01100-E2) Explain the purposes/functions/bases of the following RPS actuation circuits (Reactor Trips):

a. Manual

b. Power Range (high, low, and rate)

c. Intermediate Range

d. Source Range

e. OT(T

f. OP(T

g. Pressurizer Pressure (high and low)

h. Pressurizer Level

i. Loss of RCS Flow

j. SG Low-Low Level

k. SG Low Level w/ Feed/Steam Flow Mismatch

l. RCP UV

m. RCP UF

n. RCP Breaker Position

o. Turbine Trip (oil pressure and stop valve position)

p. Safety Injection (RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 32 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 17 ID: RO-C-EOP09-E34-1 Points: 1.00 Given the following conditions:

• A LOCA outside containment has resulted in RCS subcooling dropping to 0°F.

• The operating crew has entered OHP-4023-ECA-1.2, LOCA Outside Containment.

Which ONE of the following identifies the expected status of Containment Phase A Isolation, and the parameter used to verify that the LOCA has been isolated in accordance with ECA-1.2?

Phase A Status Parameter Monitored for LOCA Isolation A. NOT automatically actuated Pressurizer level rising B. NOT automatically actuated RCS pressure rising C. automatically actuated Pressurizer level rising D. automatically actuated RCS pressure rising Answer: D Answer Explanation:

A. INCORRECT - Pressurizer level rising is plausible since the student could reason that it may be rising if the leak was isolated. The procedure directs the use of RCS pressure increasing as the method used to indicate the leak has been isolated. Phase A not being actuated is plausible if applicant misapplies the fact that the LOCA was outside of containment, and therefore containment pressure does NOT rise to 1.0 psig. 1.0 psig is the containment pressure setpoint for Safety Injection. Since Phase A isolates certain components of containment, and since this LOCA is outside containment, it is plausible that the applicant would believe that a Phase A is not needed.

B. INCORRECT - Plausible since the procedure does direct the use of RCS pressure increasing as the method for determining the leak is isolated.

C. INCORRECT - Plausible since a Phase A has been actuated, due to the Safety Injection signal. Pressurizer level rising is plausible since the student could reason that it may be rising if the leak was isolated.

D. CORRECT - The procedure directs the use of RCS pressure increasing as the method used to indicate the leak has been isolated. A Phase A has actuated, since a Safety Injection was automatically initiated due to low RCS pressure resulting from the LOCA.

DC COOK OPS Page: 33 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-EOP09-E34-1 External Topic ID: PROC-3: PROC-3, Procedure Knowledge Level/Difficulty: H/3 Comments:

REFERENCE:

OHP-4023-E-0, pg. 22, OHP-4023-ECA-1.2, pg. 4 K/A 00WE04 EK1.1 LOCA Outside Containment Knowledge of the operational implications of the following concepts as they apply to the LOCA Outside Containment:

Components, capacity, and function of emergency systems RO - 3.5 SRO - 3.9 CFR - 41.8 / 1.10 / 45.3 K/A Match - Requires knowledge of the function of SI/Phase A Actuation that will occur and the operation of components LOCA isolations and proper operation of systems (RCS pressure rising) used during a LOCA outside of containment event.

Associated objective(s):

(RO-C-EOP09-E34) For each of the E-1 Series procedures, identify the Major Action Categories and discuss the bases for each.

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 34 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 18 ID: NRCAUDIT07-0368 Points: 1.00 The control room operators are responding to a loss of secondary heat sink.

Attempts to restore feed flow to the SG's have been unsuccessful, so the operators are attempting to establish RCS Bleed and Feed per 1-OHP-4023-FR-H-1, Loss of Secondary Heat Sink.

The need to locally align an ECCS valve has delayed the establishment of an RCS feed path for a few minutes.

Which ONE of the following describes what should be done with the PRZ PORVs while waiting for the RCS feed path to be established and why?

A. Open the PRZ PORVs first and provide RCS feed whenever the alignment is complete.

The open PRZ PORVs will provide an acceptable method of RCS heat removal until the RCS feed path is established.

B. Open the PRZ PORVs first. Opening the PRZ PORVs must not be delayed in order to comply with the procedure caution that directs RCS heat removal by Bleed and Feed to be established quickly.

C. DO NOT open the PRZ PORVs. In the event of a delay establishing Bleed and Feed, the PRZ and Reactor Vessel Head Post Accident Vents must be opened to rapidly depressurize the RCS.

D. DO NOT open the PRZ PORVs. The operator must wait until RCS feed is established before opening the PRZ PORVs because severe core damage could result.

Answer: D Answer Explanation:

A. INCORRECT - Opening the PORVs will provide some heat removal but without feed severe uncovering of the core may result.

B. INCORRECT - Opening the PORVs will provide some heat removal but without feed severe uncovering of the core may result.

C. INCORRECT - If the PORV bleed path is not adequate the vents will be used but they are not opened until feed is established.

D. CORRECT - Once the criteria is met the Feed path must be established before opening the Bleed path (PORVs) otherwise severe uncovering of the core may result.

Question ID: NRCAUDIT07-0368 Level/Difficulty: F/3 DC COOK OPS Page: 35 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Comments:

REFERENCE:

01-OHP-4023-FR-H-1, Loss of Secondary Heat Sink Steps 21, 22 (26 for Distractor)

K/A OOWE05 EK2.2 Knowledge of the interrelations between Loss of Secondary Heat Sink and the following:

Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between proper operation of these systems to the operation of the facility K/A Match Requires knowledge of the heat removal from feed and bleed operations and how this is correctly implemented during a loss of Heat sink.

Associated objective(s):

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

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 36 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 19 ID: RO-C-AOP0240412-E2-2 Points: 1.00 A reactor Startup is in progress on Unit 1. The crew has just stabilized the plant at 5% power.

The Steam Dumps are in MANUAL in Steam Pressure mode maintaining a Tave of 551.9 o F.

When the RO begins to withdraw control rods to raise reactor power, the IRPI indicator for Control Bank D1 Rod D-4 indicates 0 steps.

• MTC was -2.5 pcm / o F at these conditions.

• Control Rod D4 was determined to add 36 pcm.

Which ONE of the following describes the expected response of the reactor temperature and power?

Power will begin to lower causing temperature to also lower as:

A. MTC adds more negative reactivity due to the cooldown causing power to lower faster.

Power and Temperature will continue to lower requiring the reactor to be tripped as temperature approaches Minimum Temperature allowed for Criticality.

B. MTC adds positive reactivity due to the cooldown causing power to stabilize at the Point of Adding Heat (POAH). and temperature to stabilize ~ 547 o F.

C. MTC adds some positive reactivity due to the cooldown causing power to lower slower, but power and temperature will continue to lower requiring the reactor to be tripped as temperature approaches Minimum Temperature allowed for Criticality.

D. MTC adds more negative reactivity due to the cooldown causing power to lower faster.

Power and Temperature will continue to lower until the Point of Adding Heat (POAH).

Temperature will stabilize ~ 547 o F while the power will stabilize slightly below the POAH.

Answer: C Answer Explanation:

A. INCORRECT - MTC is listed as negative and so will add some positive reactivity. The plant will need to be tripped prior to lowering below 541 o F.

B. INCORRECT - MTC is listed as negative and so will add some positive reactivity but it will not be enough to stabilize the plant. The negative reactivity from the dropped rod will continue to lower power.

C. CORRECT - MTC is listed as negative and so will add some positive reactivity but it will not be enough to stabilize the plant. The negative reactivity from the dropped rod will continue to lower power. The plant will need to be tripped prior to lowering below 541 o F. (11 deg

• 2.5 = 27.5 pcm from MTC) since dumps are in manual.

DC COOK OPS Page: 37 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. INCORRECT - The Steam Dumps are in Manual so temperature will continue to lower as the plant goes below the point of adding heat. Power won't stabilize.

Question ID: RO-C-AOP0240412-E2-2 Level/Difficulty: H/3 Comments:

REFRENCE: RO-C-GF04 reactivity Coefficients K/A 000003 AK1.16 Dropped Control rod Knowledge of the operational impacts of the following as they apply to Dropped control rod:

MTC RO - 2.9 SRO - 3.2 CFR - 41.8 / 41.10 / 45.3 K/A Match Question requires candidate to determine impact of a dropped rod based on MTC.

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.

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 38 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 20 ID: RO-C-AOP0340412-E1-5 Points: 1.00 Seal Injection flows are currently 7.2 GPM each.

Letdown flow is 73 GPM With the Pressurizer Level selector switch in the Channel 1/2 position, Pressurizer Level transmitter NLP-151 fails in the HIGH direction.

Legend:

QRV-111, Letdown Isolation Valve QRV-112, Letdown Isolation Valve QRV-251, Charging Header Flow Control Valve Which ONE of the following statements describes the expected response / status of the letdown and charging systems?

A. Initially all orifice valves remain open, QRV-111 and QRV-112 remain open, QRV-251 throttles to minimum stop maintaining 7.2 gpm seal injection flow.

B. Initially all orifice valves remain open, QRV-111 and QRV-112 remain open, QRV-251 throttles to minimum stop also lowering seal injection flow.

C. QRV-251 throttles completely closed causing actual level to lower and QRV-112 to trip closed. All orifice valves and QRV-111 remain open. Seal Injection flow is maintained at 7.2 gpm through QRV-200.

D. QRV-251 throttles completely closed causing actual level to rapidly lower, tripping closed all orifice valves and QRV-111. QRV-112 will remain open. Seal Injection flow is at minimum since QRV-251 is closed.

Answer: B Answer Explanation:

A. INCORRECT - The orifice and isolation valves will initially remain open. QRV-251 will throttle closed to the minimum stop of ~ 42 GPM. Seal Injection flow will be maintained but will lower since the back pressure at QRV-200 is lower.

B. CORRECT - The orifice and isolation valves will initially remain open. QRV-251 will throttle closed to the minimum stop of ~ 42 GPM. Seal Injection flow will be maintained but will lower since the back pressure at QRV-200 is lower.

C. INCORRECT - QRV-251 will only throttle to a minimum flow. The lowering level will eventually cause the bistable channel to Isolate ONLY QRV-111 on actual low level.

DC COOK OPS Page: 39 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. INCORRECT - QRV-251 will only throttle to a minimum flow. The lowering level will eventually cause the bistable channel to Isolate ONLY QRV-111 on actual low level.

Question ID: RO-C-AOP0340412-E1-5 Level/Difficulty: H/3 Comments:

REFERENCE:

RO-C-00202 Pressurizer K/A 000028 AK2.03 Pressurizer (PZR) Level Control Malfunction Knowledge of the interrelations between the Pressurizer Level control Malfunctions and the following:

Controllers and Positioners RO - 2.6 SRO - 2.9 CFR - 41.7 / 45.7 K/A Match Requires knowledge of how the controllers and positioners respond to a level failure.

Associated objective(s):

(RO-C-AOP0340412-E1) Identify the event and predict the response of the plant to a Pressurizer Level Instrument Malfunction or Controller Failure, including final plant configuration.

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 40 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 21 ID: RO-C-AOP0130412-E3-1 Points: 1.00 Given the following conditions:

• The refueling crew has just set a new fuel assembly in the Unit 1 core.

• The Unit 1 core is 50% reloaded.

• The manipulator crane operator observes the refueling cavity level lowering.

• The SRO-CA and Control Room begin implementing 1-OHP-4022-002-006, Loss of Refueling Water Level during Refueling Operations.

Which one of the following describes the reason for activating the Containment evacuation alarm during this event?

The loss of refueling cavity level will:

A. require immediately stopping the RHR which can lead to core uncover within 30 minutes as the RCS boils.

B. lower the amount of shielding of the installed fuel which will increase the radiation levels.

C. require the purge fans to be stopped and all hatches closed to limit the spread of airborne iodine activity.

D. cause dangerous conditions due to shorting of electrical components from the leakage.

Answer: B Answer Explanation:

A. INCORRECT - If the level continues to lower below 614 (~ mid loop) RHR pumps may need to be stopped which could lead to boiling within 30 minutes but uncover will not occur for a couple of hours.

B. CORRECT - Since half of the core has been reinstalled, a large percentage of the Fuel assemblies will be activated which could lead to rising activity levels as the water level lowers.

C. INCORRECT - The purge fans will be stopped to aid in traveling through the hatches which were closed. The Iodine concern is from a failed fuel assembly.

D. INCORRECT - The loss of cavity level could cause water to be in undesirable areas, but most of the containment equipment is rated for a hazardous environment and this is not the main concern DC COOK OPS Page: 41 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-AOP0130412-E3-1 Level/Difficulty: F/3 Comments:

REFERENCE:

1-OHP-4022-002-006 K/A - 000036 AK3.01 Fuel Handling Incidents Knowledge of the reasons for the following responses as they apply to the Fuel Handling Incidents:

Different inputs that will cause a reactor building evacuation RO - 3.1 SRO - 3.7 CFR - 41.5 / 41.10 / 45.6 / 45.13 K/A Match Requires candidate to recognize lowering cavity level with fuel in the core will cause an increase in radiation levels in containment and containment evacuation is required to protect personnel.

Associated objective(s):

(RO-C-AOP0130412-E3) explain the procedural mitigation strategy for a Loss of Refueling Cavity Water Level (RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 42 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 22 ID: NRCAUDIT07-0011 Points: 1.00 Unit 1 has just entered mode 3 following a maintenance outage to replace a leaky fuel assembly. During work on the East RHR pump, an accidental spill causes radiation levels to increase.

The following radiation channels have alarmed:

• ERA-7305 U1 East RHR Pump Room - RED

• VRS-1505 Unit Vent Effluent Low Range Noble Gas - RED Which ONE of the following is true regarding system operations based on these conditions?

A. The Auxiliary Building Supply fans have automatically tripped.

B. The AES Fan Charcoal Filter has automatically aligned.

C. The AES Fan Charcoal Filter must be manually placed in service.

D. The Auxiliary Building Supply fans must be manually tripped.

Answer: C Answer Explanation:

A. INCORRECT - Auxiliary Building Supply fans are not automatically tripped. (Fuel Pool area fans are tripped on local radiation)

B. INCORRECT - The dampers realign for flow through the charcoal filter bed when actuation by the manual selector switch or a Phase B actuation signal occurs.

C. CORRECT - If the alarm actuates on the ERA-7300 pump rooms the Operator is required to place the AES Fan Charcoal Filter test Selector Switch to the CHAR FILT position.

D. INCORRECT - Auxiliary Building Supply are not stopped.

Question ID: NRCAUDIT07-0011 External Topic ID: 2801B: 2801B, Auxiliary Building Ventilation Level/Difficulty: F/4 Comments:

REFERENCE:

12-OHP-4024-139, Annunciator Response: Radiation Drop 11 DC COOK OPS Page: 43 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 000060 AA1.02 Accidental Gaseous Radwaste Release Ability to operate and / or monitor the following as they apply to the Accidental Gaseous Radwaste Release; Ventilation System RO - 2.9 SRO - 3.1 CFR - 41.7 / 45.5 / 45.6 K/A Match Requires ability to monitor the AES Fan Charcoal Filters and place them in service during a gaseous release.

Associated objective(s):

(RO-C-02801B-E3) Describe the function of the following Auxiliary Building Ventilation System Major Components:

a. Fans

b. Filters

c. Heaters

d. Charcoal Adsorbers

e. Backdraft Dampers

f. Volume Control Dampers

g. Remotely Operated Dampers (RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 44 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 23 ID: RO-C-AS17-E3-1 Points: 1.00 Given the following:

• A Main Transformer fire has occurred on Unit 1.

• The Unit 1 Reactor and Main Turbine/Generator have been tripped.

• The Turbine AEO has reported that the Main Transformer deluge system has actuated.

• The Outside Tour AEO has reported that all three fire water pumps are running.

The reported status of the Fire Water System is ____________ for this event. The Main Transformer deluge system ___________ expected to automatically actuate once the Main Generator is tripped, ___________ fire water pumps are expected to be running.

A. abnormal; is; but only 2 B. abnormal; is not; but 3 C. normal; is; and 3 D. abnormal; is not; and only 2 Answer: A Answer Explanation:

A. CORRECT - A main transformer fire is the largest credited fire on site. Only 2 fire pumps are required by design. Deluge is activated once the generator trips.

B. INCORRECT - Only 2 fire pumps are required by design. Deluge is activated once the generator trips.

C. INCORRECT - Only 2 fire pumps are required by design. Deluge is activated once the generator trips.

D. INCORRECT - Deluge is activated once the generator trips.

Question ID: RO-C-AS17-E3-1 External Topic ID: AS17: AS17, Fire Water Protection Comments:

REFERENCE:

1-OHP-4024-101 Drop 1 K/A 000067 2.1.28 Plant Fire on Site Conduct of operations Knowledge of the purpose and function of major component s and controls.

RO - 4.1 SRO - 4.1 DC COOK OPS Page: 45 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 CFR - 41.7 K/A Match Requires candidate to determine actions required for a fire based on the component functions.

Associated objective(s):

(RO-C-AS17-E3) Describe the operation of the main turbine fire protection system to include the following:

a. Detection.

b. Alarms.

c. Initiation.

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 46 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 24 ID: NRCAUDIT07-0045 Points: 1.00 Following a LOCA with subsequent ECCS failures, the crew is performing the actions in 01-OHP-4023-FR-C.2, Response to Degraded Core Cooling.

The following conditions exist:

• RCS Pressure - RISING

• Annunciator #108 Drop 6 Pressurizer Pressure High - High - LIT

• Pressurizer PORVs are open

• Core Cooling has NOT been restored Which ONE of the following describes the required operation of the Pressurizer PORVs in this event?

A. Allow the PORVs to operate for RCS overpressure control as necessary.

B. Place the PORVs in OPEN to depressurize the RCS until the accumulators inject.

C. Isolate the PORVs to prevent further loss of RCS inventory.

D. Place the PORVs in CLOSE until required for inadequate core cooling conditions.

Answer: A Answer Explanation:

A. CORRECT - Per Step 6 of 01-OHP-4023-FR-C.2, the PORVs are verified closed with block valves open. This is to ensure that a PORV remains available to preclude use of the safety valves. Given the High pressure alarm (2385 psig) the PORVs should be allowed opened to reduce pressure and then verified reclosed.

B. INCORRECT - The RCS is not depressurized with the PORV in FR-C.2 (although this may be performed in FR-C.1)

C. INCORRECT - The PORV should not be isolated since its operation is required to prevent lifting the safety valves.

D. INCORRECT - The PORV should not be placed in close since its operation is required to prevent lifting the safety valves.

DC COOK OPS Page: 47 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0045 Comments:

REFERENCE:

01-OHP-4023-FR-C.2, Response to Degraded Core Cooling 12-OHP-4023-FR-C.2 Background Document K/A 00WE06 EA2.2 Degraded Core Cooling Ability to interpret the following as they apply to the Degraded Core Cooling:

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments RO-3.5 SRO-4.1 CFR - 41.7 / 41.10 K/A Match Requires candidate to determine status of PORVs which are operated to limit pressure during a degraded Core Cooling condition.

Associated objective(s):

(RO-C-EOP10-E12) For each of the FR-C and FR-I series procedures identify the Major Action Categories and discuss the bases for each.

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 48 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 25 ID: RO-C-EOP-E10-2 Points: 1.00 Given the following conditions:

The crew is responding to Reactor Trip following a Loss of Control Air.

The Unit Supervisor entered 1-OHP-4023-FR-H.2, Response to Steam Generator Overpressure to address the high pressure on SG #11.

S/G # Level (% NR) Pressure (PSIG) 11 60 1200 12 40 1005 13 40 1005 14 40 1005 Feedwater Isolation has been verified and AFW was isolated to SG 11 as directed by 1-OHP-4023-FR-H.2.

The procedure also directs that AFW remain isolated to SG #11 until a steam relief path is established.

Which one of the following describes the reason for maintaining AFW isolated?

A. Operation of the TDAFP at high steam pressure may cause turbine damage.

B. Adding cold AFW can cause additional flashing on the hot tubes, causing pressure to rise even higher.

C. Differential temperatures between the existing high pressure conditions in the SG and the cold AFW will cause tube failure due to thermal stress.

D. AFW is a high pressure system which may cause further overpressure in a SG with no steam release path.

Answer: D Answer Explanation:

A. INCORRECT - Plausible because the TDAFP is normally in service at lower steam pressure.

B. INCORRECT - Plausible because the same type of feed restrictions exist in both FR-H.1 and FR-H.5 with the concern of tube failure when feeding a hot SG. While the concern is primarily thermal stress, flashing in a dry SG could result in a pressure rise.

C. INCORRECT - Plausible because the same type of feed restrictions exist in both FR-H.1 and FR-H.5, but this is not the basis for this caution in FR-H.2.

D. CORRECT - Per the Background document, if AFW flow is reestablished to an affected SIG prior to establishing a steam release path, since AFW is a high pressure system, the AFW flow could further increase SG pressure.

DC COOK OPS Page: 49 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-EOP-E10-2 Level/Difficulty: F/3 Comments:

REFERENCE:

12-OHP-4023-FR-H-2 Background Steps 6/7 K/A 00WE13 EA1.3 Steam Generator Overpressure Ability to operate the following as they apply to the Steam generator Overpressure:

Desired operating results during normal and emergency situations RO -3.1 SRO - 3.4 CFR -41.7 / 45.5 / 45.6 K/A Match:

Requires candidate to determine correct result desired from the isolation of AFW.

Associated objective(s):

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

(RO31 NRC Exam 2018 1-25)

DC COOK OPS Page: 50 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 26 ID: RO-C-EOP09-E19-1 Points: 1.00 Given the following conditions:

• A Small Break Loss of Coolant Accident has occurred on Unit 1.

• Conditions to start a Reactor Coolant Pump cannot be established.

• Containment pressure is 5.1 psig and slowly rising.

• Reactor Coolant System (RCS) pressure is 950 psig and stable.

• All Steam Generator (SG) narrow range levels are approximately 45% and stable.

• All RCS Hot Leg temperatures are approximately 540°F and stable.

• Core Exit Thermocouples are reading approximately 550°F and rising.

• Steam Generator pressures are as follows:

SG #11 is 785 psig and stable.

SG #12 is 790 psig and stable.

SG #13 is 850 psig and stable.

SG #14 is 780 psig and stable.

• RCS Cold Leg temperatures are as follows:

Loop 1 is 518°F and stable.

Loop 2 is 519°F and stable.

Loop 3 is 360°F and lowering.

Loop 4 is 517°F and stable.

What is the status of natural circulation and the expected operator action?

A. Adequate natural circulation does NOT exist and the Steam Dump Valves should be opened farther.

B. Adequate natural circulation does exist and the PORV for SG #13 should be closed farther.

C. Adequate natural circulation does NOT exist and the SG PORVs should be opened farther.

D. Adequate natural circulation does exist and Steam Dump Valves should be closed farther.

Answer: C DC COOK OPS Page: 51 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Answer Explanation:

A. INCORRECT - Plausible as Natural Circulation does NOT exist, however, the MSIVs would have closed at 2.8 psig Containment pressure and thus the Steam Dump Valves are isolated.

B. INCORRECT - Plausible as three of the four Steam Generators appear to be coupled.

However, adequate subcooling does not exist and one Steam Generator is uncoupled. Additionally, the CETs are increasing which indicates that adequate Natural Circulation does NOT exist. SG #13 appears to be overcooling, thus the concept to decrease steam dumping from this generator is plausible however, the low temperature indicates lack of circulation and Cold Leg temperature lowering as a result of ECCS flow.

C. CORRECT - Since adequate subcooling does NOT exist, CETs are increasing and SG

1. 13 is uncoupled, which indicate that Natural Circulation does NOT exist.

SUP-011 states to increase dumping steam to promote Natural Circulation.

The Steam Dump Valves are isolated by the MSIVs which only leave the SG PORVs for dumping steam.

D. INCORRECT - Plausible as three of the four Steam Generators appear to be coupled.

However, adequate subcooling does not exist and one Steam Generator is uncoupled. Additionally, the CETs are increasing which indicates that adequate Natural Circulation does NOT exist. SG #13 appears to be overcooling, thus the concept to decrease steam dumping from this generator is plausible however, the low temperature indicates lack of circulation and Cold Leg temperature lowering as a result of ECCS flow.

Question ID: RO-C-EOP09-E19-1 Level/Difficulty: H/4 Comments:

REFERENCE:

OHP-4023-SUP-011, Steam Tables K/A 00WE03 EK2.1 LOCA Cooldown and Depressurization Knowledge of the interrelationships between the LOCA Cooldown and Depressurization and the following:

Components and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

RO - 3.6 SRO - 4.0 CFR - 41.7 / 45.7 K/A Match Candidate must evaluate plant conditions post LOCA and decide control means to be used to cooldown for depressurization.

DC COOK OPS Page: 52 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Associated objective(s):

(RO-C-EOP09-E19) Discuss Natural circulation during a SBLOCA to include indications of and factors affecting natural circulation.

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 53 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 27 ID: CM-0726 Points: 1.00 During a SGTR with a Loss of Offsite Power (No RCPs), it is possible to get an indication of a Pressurized Thermal Stress (PTS) condition. The indication comes about because of operator action to mitigate the SGTR.

Which ONE of the following is the cause of the PTS concern?

A. Isolation of the affected SG, loss of natural circulation in the affected loop resulting in pooling of cold ECCS injection flow.

B. Isolation of the affected SG, loss of natural circulation in the affected loop, lowered ECCS flow when SI is terminated.

C. Continued ECCS flow in combination with the maximum rate cooldown to the required target temperature overcools the unaffected loops.

D. Saturated conditions at the core exit during depressurization, loss of natural circulation in all loops resulting in pooling of cold ECCS injection flow.

Answer: A Answer Explanation A. CORRECT - Once the affected S/G is isolated, loss of natural circulation in the affected loop resulting in pooling of cold ECCS injection flow.

B. INCORRECT - Lowered ECCS flow is not the concern for PTS.

C. INCORRECT - Once the affected S/G is isolated, loss of natural circulation in the affected loop resulting in pooling of cold ECCS injection flow.

D. INCORRECT - Once the affected S/G is isolated, loss of natural circulation in the affected loop resulting in pooling of cold ECCS injection flow.

Question ID: CM-0726 External Topic ID: MCD: MCD, Mitigating Core Damage Comments:

REFERENCE:

12-OHP-4023-E-3 Caution prior to Step 7.

DC COOK OPS Page: 54 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 00WE08 EA2.1 Pressurized Thermal Shock Ability to determine and interpret the following as they apply to the Pressurized thermal Shock:

Facility conditions and selection of appropriate procedure during abnormal and emergency conditions.

RO - 3.4 SRO - 4.2 CFR - 41.7 / 41.10 / 43.5 / 45.13 K/A Match Candidate must interpret indications during a SGTR which may present a PTS concern to ensure the correct procedure is implemented.

Associated objective(s):

(RO-C-EOP08-E9) Explain why Pressurized Thermal Shock indications presented during a SGTR without RCPs operating are invalid and when they become valid.

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 55 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 28 ID: RO-C-00201-E3-1 Points: 1.00 A spurious Phase 'A' Containment Isolation has occurred.

Which ONE of the following describes the effect of this isolation on the Reactor Coolant Pumps (RCP's)?

A. RCP seal leakoff will be isolated. The seal leakoff flow will be directed to the Pressurizer Relief Tank (PRT) via a relief valve.

B. CCW to the RCP thermal barriers will be isolated. RCP seal injection must be maintained.

C. CCW to the RCP motor coolers will be isolated. The RCPs will have to be manually tripped.

D. RCP seal injection is isolated. CCW to the thermal barriers must be maintained.

Answer: A Answer Explanation:

A. CORRECT - Phase A closes RCP Seal Water Returns QCM-250 and QCM-350. The upstream relief valve opens at 150# to allow seal return water to the PRT.

B. INCORRECT - CCW flow would be isolated by a Phase B isolation signal.

C. INCORRECT - CCW flow would be isolated by a Phase B isolation signal.

D. INCORRECT - Seal injection flow is not affected by Phase A.

Question ID: RO-C-00201-E3-1 External Topic ID: 0201: 0201, Reactor Coolant Pump Level/Difficulty: F/2 Comments:

REFERENCE:

RO-C-00201, RO-C-00300, OHP-4023-SUP-003 K/A 003000 K6.04 Reactor Coolant Pump System (RCPS)

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

Containment isolation valves affecting RCP operation RO - 2.8 SRO - 3.1 CFR - 41.7 / 45.5 DC COOK OPS Page: 56 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A Match This meets the KA of knowledge of the effect of a malfunction of containment isolation valves on RCP operation. The RCPs themselves may continue to operate in this situation, but the operation of the RCP's seal return flow has been affected by the spurious Phase A isolation.

Associated objective(s):

(RO-C-00201-E3) Describe the seal injection flowpath and flow distribution for the RCP seal package.

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 57 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 29 ID: CM-39317A Points: 1.00 Unit 1 was operating at 100% power. The West CCP had been tagged out due to a bearing failure.

The following sequence of events occurs:

• A reactor trip and Safety Injection occurs in response to a Steam Generator Tube Rupture.

• Safety Injection is NOT reset.

• The T11D, 4kV AC ESF Bus subsequently loses normal power but is re-energized by the 1CD Emergency Diesel Generator.

• T11D Automatic load sequencing is complete.

As a result of these events, the East CCP:

A. has tripped and automatically restarted.

B. was NOT affected by the loss of Bus T11D.

C. has tripped and must be manually started.

D. has tripped and may NOT be manually started until the load conservation signal resets.

Answer: A Answer Explanation:

A. CORRECT - Since the SI signal was still active the pump sheds and then starts when the EDG energizes the bus.

B. INCORRECT - Plausible if the East and West power supplies are mistakenly swapped.

Incorrect as the East CCP is powered from T11D.

C. INCORRECT - The ECCS timer and SI signals are the only auto starts for the CCPs. The pumps will automatically start following a load shed since SI is NOT reset.

The East CCP is powered from T11D while the West CCP is powered from T11A. This answer would be correct if SI was reset.

D. INCORRECT - Plausible as it could be thought that load conservation signal would prevent a CCP start. Incorrect as the CCP is not prevented from starting due to a load conservation.

DC COOK OPS Page: 58 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: CM-39317A External Topic ID: 0300-1: 0300-1, CVCS Level/Difficulty: H/3 Comments:

REFERENCE:

12-OHP-4023-E-0 Step 26 Caution Modified CM-39317 by changing Stem to SI has NOT Been Reset which changes correct Answer to A from C K/A 004000 K2.03 Chemical and Volume Control system (CVCS)

Knowledge of bus power supplies to the following:

Charging pumps.

RO - 3.3 SRO - 3.5 CFR - 41.7 K/A Match Candidate must determine operation of the CCP based on signals and power supplies.

Associated objective(s):

(RO-C-00300-E14) Given the following CVCS components, describe the conditions that will cause the component to trip, automatically/manually start and/or automatically/manually reposition (including any control interlocks).

a. Letdown Isolation Valves

b. Letdown Orifice Isolation Valves

c. Demineralizer Divert Valve

d. VCT Divert Valve

e. Charging Pumps (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 59 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 30 ID: RO-C-00300-E12-1 Points: 1.00 Given the following plant conditions:

Unit 2 is operating at 100%.

120 GPM letdown is in service.

Volume Control Tank (VCT) blender controls are in automatic.

VCT level transmitter QLC-452 fails to 100%.

Which ONE of the following describes the expected impact to the VCT?

With NO operator action, the VCT level will lower:

A. slowly due to normal RCS losses. Auto VCT makeup will NOT start and eventually the VCT level will lower to the RWST switchover setpoint. With QLC-452 failed high, the coincidence for switchover will not be made up and level will continue lowering to 0%.

B. due to Letdown diverting. Auto VCT makeup will NOT start and eventually the VCT level will lower to the RWST switchover setpoint. QLC-451 will actuate a switchover to RWST Suction on Lo-Lo VCT level.

C. due to Letdown diverting. Auto VCT makeup will attempt to control VCT level. Eventually VCT level will lower to the RWST switchover setpoint. With QLC-452 failed high, the coincidence for switchover will not be made up and level will continue lowering to 0%.

D. slowly due to normal RCS losses. Auto VCT makeup will attempt to control VCT.

Eventually VCT level will lower to the RWST switchover setpoint. QLC-451 will actuate a switchover to RWST Suction on Lo-Lo VCT level.

Answer: C Answer Explanation:

A. INCORRECT - QLC-452 will cause full divert and does not impact makeup.

B. INCORRECT - QLC-452 does not impact makeup and switchover to the RWST requires both channels.

C. CORRECT - QLC-452 causes full divert, QLC-451 actuates makeup and switchover to the RWST requires both channels.

D. INCORRECT - QLC-452 causes full divert and switchover to the RWST requires both channels.

DC COOK OPS Page: 60 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-00300-E12-1 External Topic ID: 0300-2: 0300-2, CVCS Makeup System Level/Difficulty: H/3 Comments:

REFERENCE:

SOD-00300-002, 2-OHP-2024-209 Drop 50 K/A 004000 A1.06 Chemical and Volume Control System (CVCS)

Ability to predict and / or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including:

VCT level RO - 3.0 SRO - 3.2 CFR - 41.5 / 45.5 K/A Match This question requires knowledge of the required logic for lo-lo VCT level and the response of the system to a level channel failure.

Associated objective(s):

(RO-C-00300-E12) Describe the automatic actions and setpoints for the VCT level control system.

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 61 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 31 ID: NRCAUDIT07-0687 Points: 1.00 The following conditions exist:

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

• West RHR Pump is operating with the return aligned to the Alternate Cooldown Path to Loops 2 & 3.

• RCS temperature is 300°F and stable.

• RCS pressure is 335 psig and stable.

The air supply line to IRV-320, West RHR Hx Outlet Valve, breaks, causing a complete loss of Instrument Air to the valve.

Which ONE (1) of the following describes the effect on the plant and the action that could be taken to mitigate the transient?

A. RHR Flow through the West HX will be lost. Throttle open IRV-311 RHR HX Bypass to maintain greater than 3000 gpm RHR flow.

B. RHR Flow through the West HX will be lost. Stop the West RHR pump immediately to prevent over pressurizing letdown.

C. RHR Flow through the West HX will rise. Throttle ICM-111, RHR Discharge to Cold Leg 2

& 3 and IRV-311 RHR HX Bypass to prevent overcooling the RCS.

D. RHR Flow through the West HX will rise. Throttle ICM -321, West RHR Injection to Loops 2 & 3 and IRV-311 RHR HX Bypass to prevent overcooling the RCS.

Answer: D Answer Explanation:

A. INCORRECT - IRV-320 fails open so flow will rise. Plausible since RHR flow is maintained >

3000 gpm to minimize vibrations & cavitation through the piping.

B. INCORRECT - IRV-320 fails open so flow will rise. Plausible since RHR flow to the letdown system taps off before the IRV-320 and if IRV-320 closed it may raise letdown pressure.

C. INCORRECT - The ICM-111, RHR Discharge to Cold Leg 2 & 3 is in the NORMAL Cooldown Path and should only be used if the Alternate cooldown path is unavailable.

This would be correct if the Normal Cooldown path was used.

D. CORRECT - IRV-320 fails open on loss of air. This will raise RHR flow through the HX.

The ICM-321 can be throttled closed to reduce total RHR flow and IRV-311 can be throttled open to allow more flow to bypass the HX in order to control RCS cooldown.

DC COOK OPS Page: 62 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0687 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)

K/A 005000 A2.04 Residual Heat Removal System (RHRS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

RHR valve malfunction RO - 2.9 SRO - 2.9 CFR - 41.5 / 43.5 / 45.3 / 45.13 K/A Match Candidate is required to determine impact of RHR valve malfunction and actions to be taken to combat malfunction.

Associated objective(s):

(RO-C-01700-E4) Given the following support system(s), 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 (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 63 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 32 ID: RO-C-01700-E7-10 Points: 1.00 Which ONE of the following conditions describes the operation of the Containment Recirculation Sump Valve, ICM-305 Control Power Switch and indicating Lights?

The Control Power Switch must be Placed to ON to enable the:

A. valve to be opened OR closed. The Center Cube is illuminated when the switch is placed to ON. The Outer Lights will Illuminate when the interlocks to open the valve are met.

B. valve to be opened OR closed. The Center Cube is illuminated whenever the valve breaker has power. The Outer Lights will Illuminate when the switch is placed to ON.

C. valve to be CLOSED ONLY. The Center Cube is illuminated whenever the valve breaker has power. The Outer Lights will Illuminate when the switch is placed to ON.

D. valve to be OPENED ONLY. The Center Cube is illuminated when the switch is placed to ON. The Outer Lights will Illuminate when the interlocks to open the valve are met.

Answer: A Answer Explanation:

A. CORRECT - Center Cube is lit with switch on and outer lights will illuminate when interlocks are made up. Switch must be on for Open or Close.

B. INCORRECT - Center cube is normally dark until switch is turned on. Outer lights illuminate only when interlock is made up.

C. INCORRECT - Center cube is normally dark until switch is turned on. Outer lights illuminate only when interlock is made up.

D. INCORRECT - Center Cube is lit with switch on and outer lights will illuminate when interlocks are made up. Switch must be on for Open AND Close.

Question ID: RO-C-01700-E7-10 Level/Difficulty: F/4 Comments:

REFERENCE:

12-OHP-4023-ES-1.3, OP-1-982841 DC COOK OPS Page: 64 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 006000 K4.18 Emergency Core Cooling System (ECCS)

Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following:

Valves normally isolated from their control power RO - 3.6 SRO - 3.7 CFR - 41.7 K/A Match Candidate is required to demonstrate knowledge on control power for valves that are normally isolated and indications when it is applied to them.

Associated objective(s):

(RO-C-01700-E7) Explain the control interlocks associated with the operation of the following RHR System components.

a. IMO-330/331

b. IMO-340/350

c. IMO-128/ICM-129

d. ICM-305/306

e. IMO-310/320 (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 65 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 33 ID: NRCAUDIT07-0003 Points: 1.00 The plant has experienced a large break LOCA. The reactor has tripped and an SI signal is present. Charging Pumps Suction from RWST, IMO-910 opened BUT IMO-911 failed to open.

Which ONE of the following is the impact of this action on the Charging pump suction flow?

A. Both operating Charging pumps will receive suction flow from the VCT.

B. Both operating Charging pumps will receive suction flow from the RWST.

C. One Charging pump will continue to take suction from the VCT and one Charging pump will take suction from the RWST.

D. One Charging pump will take suction from the RWST. The other Charging pump will have no suction until operator action is taken.

Answer: B Answer Explanation:

A. INCORRECT - The VCT suction valves will isolate on a SI signal after IMO-910 and/or IMO911 are OPEN. Either valve will allow VCT Suction Valves to Close.

B. CORRECT - CVCS suction from the RWST is provided through the IMO-910 and IMO-911 valves which are in a parallel arrangement allowing either valve to provide suction to both pumps.

C. INCORRECT - The VCT suction valves will isolate on a SI signal after IMO-910 and/or IMO911 are OPEN. Either valve will allow VCT Suction Valves to Close.

D. INCORRECT - IMO-910 and IMO-911 valves which are in a parallel arrangement allowing either valve to provide suction to both pumps.

DC COOK OPS Page: 66 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0003 External Topic ID: 0300-1: CVCS Comments:

REFERENCE:

RO-C-00300, Chemical Volume Control System pg. 47 SOD-00300-001 Charging and Letdown System drawing K/A 006000 A3.08 Emergency Core Cooling system (ECCS)

Ability to monitor automatic operation of the ECCS including:

Automatic transfer of ECCS flow paths RO-4.2 SRO-4.3 CFR - 41.7 / 45.5 K/A Match Candidate is required to monitor (and determine if acceptable) automatic actions (or failure of) on 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.

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 67 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 34 ID: NRCAUDIT07-0393 Points: 1.00 Unit 2 is operating at 50% power.

The following conditions exist:

• Pressurizer pressure is 2235 psig

• Pressurizer Relief Tank (PRT) pressure is 20 psig

• PRT temperature is 125°F

• PRT level is 81%

• The PRT is being cooled by spraying from the Primary Water

• A pressurizer code safety valve is suspected of leaking by its seat What temperature would be indicated on the safety valve tailpipe temperature if the code safety were leaking by?

A. 227-231°F B. 258-262°F C. 550-552°F D. 652-654°F Answer: B Answer Explanation:

A. INCORRECT - Temp is too low - the correct temp is 260°F Plausible: If the candidate makes the mistake of not correcting for atmospheric pressure by failing to adding 14.6 psi to the PRT pressure and uses 20 psia.

B. CORRECT - The correct temp is 260°F. Steam pressure of ~2250 psia (2235 psig) constant enthalpy to ~35 psia (20 psig).

C. INCORRECT - Temp too high - Plausible if candidate thinks temperature is the same as no-load Tave.

D. INCORRECT - Temp too high - Plausible if candidate thinks temperature is the same as the Pressurizer vapor temp.

DC COOK OPS Page: 68 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0393 Comments:

REFERENCE:

Steam Tables REFERENCE PROVIDED: Steam Tables K/A 007000 A4.10 Pressurizer Relief Tank / Quench Tank system (PRTS)

Ability to manually operate or monitor in the control room:

Recognition of leaking PORV / Code Safety K/A Match Candidate is required to monitor the temperature and determine the expected indication based on thermodynamic properties.

Associated objective(s):

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 69 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 35 ID: RO27AUDIT-35 Points: 1.00 Which ONE of the following conditions would result in implementation of OHP-4023-E-0, Reactor Trip or Safety Injection?

A. Thermal barrier heat exchanger flow path isolated due to high differential flow.

B. The level on the East side of the CCW Surge tank drops to 40 inches and stabilizes.

C. ESW flow is lost to the West CCW Heat Exchanger.

D. Loss of the East CCW pump while the West CCW pump is tagged out of service.

Answer: D Answer Explanation:

A. INCORRECT - Plausible, since the isolation affects RCP parameters, but with seal injection still available, the plant can remain at power.

B. INCORRECT - Plausible, because the A side of the CCW surge tank dropping low enough to cause signs of loss of suction to CCW Pumps would require tripping all only the affected CCW pump.

C. INCORRECT - Plausible as ESW cools CCW, a loss of ESW flow to one CCW heat exchanger does NOT require the crew to trip the reactor.

D. CORRECT - Loss of ALL CCW pumps requires a reactor trip.

Question ID: RO27AUDIT-35 Level/Difficulty: F/2 Comments:

REFERENCE:

2-OHP-4022-016-004, Loss of Component Cooling Water 2-OHP-4022-016-001, Malfunction of the CCW System K/A 008000 2.1.20 Component Cooling Water System (CCWS)

Ability to interpret and execute procedural steps.

RO - 4.6 SRO - 4.6 CFR - 41.10 / 43.5 / 45.12 K/A Match Candidate is required to determine which set of conditions requires the execution of the procedural step to trip the reactor and enter E-0.

DC COOK OPS Page: 70 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Associated objective(s):

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 71 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 36 ID: 2008NRC-0620-SRO Points: 1.00 Given the following plant conditions on Unit 2:

• Reactor power is 12%.

• The controlling Pressurizer (PRZ) Pressure Channel slowly fails high.

• The RO takes manual control of PRZ Pressure Master Controller and returns demand to normal.

• All PRZ heaters have energized.

• RCS pressure is 2075 psig and slowly lowering.

• You notice that NRV-163 (PRZ spray) is OPEN.

• When the RO places NRV-163 in manual it will NOT close.

Which ONE of the following is the proper sequence of actions to stop the pressure reduction?

A. Manually trip the reactor.

Go to OHP-4023-E-0, Reactor Trip Or Safety Injection.

Trip RCP #3.

Dispatch an AEO to locally isolate Spray Valve NRV-163.

B. Reduce Power to 8%.

Trip RCPs #3 and #4.

Go to 1-OHP-4021-001-003 Power Reduction Dispatch an AEO to locally isolate Spray Valve NRV-163.

C. Trip RCP #3.

Go to OHP-4023-E-0, Reactor Trip Or Safety Injection.

D. Manually trip the reactor.

Go to OHP-4023-E-0, Reactor Trip Or Safety Injection.

Trip RCPs #3 and #4.

Answer: D Answer Explanation:

A. INCORRECT - Plausible since a Reactor Trip is necessary and isolating NRV-163 would theoretically stop the pressure lowering. Incorrect as NRV-163 does not have a manual isolation valve.

B. INCORRECT - Plausible since being less than 10% power blocks loop loss of flow trip.

Incorrect since Cook plant is not licensed to operate at power without all RCPs running C. INCORRECT - Plausible since a reactor trip and RCP trip is required. Per operating practices, the Reactor is tripped first and then the RCPs.

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. CORRECT - Three loop operations is not allowed. The RCP #3 & 4 must be stopped to stop the spray flow. Therefore the Reactor must be manually tripped and then the RCPs tripped.

Question ID: 2008NRC-0620-SRO External Topic ID: 0200-1: Reactor Coolant System - Mechanical Level/Difficulty: H/3 Comments:

REFERENCE:

2-OHP-4024-207, Annunciator #207 Drop 61, 2-OHP-4023-ES-0-1, Reactor Trip Response step 5 K/A 010000 A2.02 Pressurizer Pressure Control System (PZR PCS)

Ability to (a) predict the impacts of the following malfunctions or operations on the PZR PCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Spray valve failures RO - 3.9 SRO - 3.9 CFR - 41.5 / 43.5 / 45.3 / 45.13 K/A Match Candidate is required to determine impact of failed open spray valve and the required actions to stabilize the plant.

Associated objective(s):

(RO-C-AOP0330412-E2) explain the required operator actions to stabilize plant conditions after a Pressurizer Pressure Instrument Malfunction or Controller Failure prior to formal procedure implementation (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 73 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 37 ID: RO-C-00202-E6-1 Points: 1.00 Given the following plant conditions:

• Unit 2 is at 100% power, with all systems in normal alignment.

• NPP-151, Pressurizer Pressure Channel, failed to 2185 psig (controlling channel).

The INITIAL response of the pressure control system is the __(1)__. AFTER an OPERABLE pressure channel is selected, RCS pressure will return to __(2)__.

___(1)___ ___(2)___

A. spray valves and a PORV will be OPEN 2085 psig B. spray valves and a PORV will be OPEN 2235 psig C. backup and variable heaters will be full ON 2085 psig D. backup and variable heaters will be full ON 2235 psig Answer: D Answer Explanation:

A. INCORRECT - Spray valves will not open as pressure has lowered from the Unit 2 design value of 2235 psig causing the heaters to energize. 2085 psig is the previous Unit 1 design pressure.

B. INCORRECT - Spray valves will not open as pressure has lowered from the Unit 2 design value of 2235 psig causing the heaters to energize. 2235 psig is the correct pressure that will be restored.

C. INCORRECT - Pressure has lowered from the Unit 2 design value of 2235 psig, causing the heaters to energize. 2085 psig is the previous Unit 1 design pressure.

D. CORRECT - Pressure has lowered from the Unit 2 design value of 2235 psig, causing the heaters to energize. 2235 psig is the correct pressure that will be restored.

Question ID: RO-C-00202-E6-1 External Topic ID: 0202-1: Pressurizer/Pressure Relief Level/Difficulty: H/2 Comments:

REFERENCE:

RO-C-00202 TP71 DC COOK OPS Page: 74 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 010000 A4.02 Pressurizer Pressure Control (PZR PCS)

Ability to manually operate and / or monitor in the control room:

PZR heaters RO - 3.6 SRO - 3.4 CFR - 41.7 / 45.5 / 45.6 / 45.7 / 45.8 K/A Match Candidate is required to monitor Pressurizer Heaters to control RCS pressure.

Associated objective(s):

(RO-C-00202-E6) Describe the control signal flowpath for the Pressurizer Pressure Control System, starting with the control inputs and passing sequentially through the following components until the final control or protection output is developed:

a. Pressurizer Pressure Channels

b. Master Pressure Controller

c. Pressurizer Spray Valve Controllers

d. Spray Valves

e. Pressurizer Heaters

f. Pressurizer Power Operated Relief Valves (PORVs)

g. Reactor Protection System (RPS)/SI (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 75 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 38 ID: CM-3007A Points: 1.00 A Unit 1 Safety Injection will occur when pressurizer pressure lowers to ___(1)___ psig (actuation setpoint) in ___(2)____ channels unless the actuation has been blocked when permissive ___(3)____ became active.

1 2 3 A. 1950 two of four TAVG LO-LO PERM P-12 B. 1775 two of four TAVG LO-LO PERM P-12 C. 1950 two of three PRESSURE PERM P-11 D. 1775 two of three PRESSURE PERM P-11 Answer: D Answer Explanation:

A. INCORRECT - Wrong setpoint (U1 Lo PRZ Press. Trip), wrong coincidence (PRZ has 4 press. channels), wrong permissive (P-12 Blocks Steamline SI).

B. INCORRECT - Correct SI Setpoint, wrong coincidence, wrong permissive.

C. INCORRECT - Wrong setpoint, correct coincidence, correct permissive.

D. CORRECT - Correct setpoint, correct coincidence, correct permissive.

Question ID: CM-3007A External Topic ID: 1100-1: RPS/ESFAS Signal Level/Difficulty: F/2 Comments:

REFERENCE:

RQ-C-KNOW, RO-C-01100 K/A 012000 A3.06 Reactor Protection System Ability to monitor automatic operation of the RPS, including:

Trip logic RO - 3.7 SRO - 3.7 CFR - 41.7 / 45.5 DC COOK OPS Page: 76 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A Match Candidate is required to have knowledge of the trip logic as part of monitoring automatic operation of the RPS.

Associated objective(s):

(RO-C-01100-E6) For any RPS/ESFAS/AMSAC actuation circuit:

a. State the total number and type of process sensor inputs.

b. State the logic/coincidence of signals needed to actuate the protection system.

c. State the nominal setpoint.

d. Describe any blocking signals/permissive/conditionals.

e. Describe any associated status lights.

f. Describe any associated alarms.

g. Explain what plant actions occur in direct response to the actuation signal.

h. Explain how the sensor is "tripped" or "bypassed" to support T.S. action statement or surveillance requirements.

i. Explain when "bypass" is allowed to be used per T.S. 3.3.1-1 and 3.3.2-2.

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 77 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 39 ID: 2008NRC-0429 Points: 1.00 Given the following plant conditions:

• Unit 2 is at 100% power.

• Containment pressure instrument Channel #3, 2-PPP-301, declared inoperable.

• Required actions per 2-OHP-4022-013-011 Containment Instrumentation Malfunction have been completed.

CRID 2 has just lost power.

Which ONE of the following statements explains the impact on the Safety Injection system and expected operator actions?

A. No Safety Injection logic is satisfied. Implement 2-OHP-4021-082-008, Operation of CRID Power Supplies, to address the CRID Failure.

B. Only Train "A" Safety Injection actuation logic was satisfied, but equipment will NOT automatically actuate. Implement 2-OHP-4023-E-0, Reactor Trip or Safety Injection, and inform the operators that they must individually reposition/start the equipment.

C. Train "A" and "B" Safety Injection actuation logic were satisfied. Implement 2-OHP-4023-E-0, Reactor Trip or Safety Injection. Both Trains equipment will automatically actuate.

D. Train "A" and "B" Safety Injection actuation logic were satisfied, but Train "A" equipment will NOT automatically actuate. Implement 2-OHP-4023-E-0, Reactor Trip or Safety Injection, and inform the operators that they must individually reposition/start each Train "A" component.

Answer: C Answer Explanation:

A. INCORRECT - SI will Actuate.

B. INCORRECT - Both Trains of SI will Actuate (All channels input into both trains)

C. CORRECT - Channels 2, 3, & 4 of Containment Pressure cause an SI. This signal will result from the bi-stables being tripped on Channel 3 & the Loss of Power to Channel 2 bi-stables. The SI signal is actuated on Both SI trains. Train A SSPS output is powered from CRID 1 & Train B is powered from CRID 4 so all actuations will occur automatically.

D. INCORRECT - Both Trains output relays still have power.

DC COOK OPS Page: 78 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: 2008NRC-0429 Level/Difficulty: H/3 Comments:

REFERENCE:

RO-C-01100 Slide 153; 02-OHP-4022-013-011; 02-OHP-4021-082-008 K/A 013000 A2.04 Engineered Safety Features Actuation System (ESFAS)

Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss if Instrument bus.

RO - 3.6 SRO - 4.2 CFR - 41.5 / 43.5 / 45.3 / 45.13 K/A Match Candidate is required to demonstrate knowledge on redundancy of inputs into the ESFAS system and the ability to determine if coincidence is made up for actuation of the system and then use procedure steps / actions to mitigate the consequences.

Associated objective(s):

(RO-C-AOP0350412-E2) explain the required operator actions to stabilize plant conditions after a Containment Pressure Instrument Malfunction prior to formal procedure implementation (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 79 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 40 ID: RO-C-02800-E9-3 Points: 1.00 Unit 1 has experienced degraded voltage which led to the loss of all 4 T buses.

The Reactor tripped and a Pressurizer Safety stuck partially open leading to an SI.

The following conditions exist:

• The Safety Injection and Phase A Signals have NOT been reset.

• Emergency Diesel Generators have started and loaded as designed.

• The remaining buses (RCP Buses) have recovered to their normal voltages.

Which ONE of the following describes the status, and the required actions, if any to restart the Containment Ventilation fans?

A. The Lower and Upper Ventilation Fans remain running since they are NOT powered through the EDGs.

B. The Lower Ventilation Fans remain running since they are NOT powered through the EDGs.

The Upper Ventilation Fans have tripped on the Phase A Signal.

The Upper Ventilation fans will start once Phase A is reset.

C. The Lower and Upper Ventilation Fans have tripped on the Phase A Signal and are locked out on the load shed.

The Lower and Upper Ventilation Fans may be restarted by resetting Phase A and the load Shed after 75 seconds.

D. The Lower Ventilation Fans have restarted after their bus was powered through the EDGs.

The Upper Ventilation Fans have tripped on the Phase A Signal and are locked out on the load shed.

The Upper Ventilation Fans may be restarted by resetting Phase A and the load Shed after 75 seconds.

Answer: D Answer Explanation:

A. INCORRECT - The Upper and Lower Ventilation Units are powered from the T Buses through the EZC buses. The Upper Vent Units trip on a Phase A. Plausible since these fans are not routinely used for accident conditions.

B. INCORRECT - The Upper and Lower Ventilation Units are powered from the T Buses through the EZC buses. The Upper Vent Units trip on a Phase A but also will be locked out with load shed. Plausible since these fans are not routinely used for accident conditions.

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 C. INCORRECT - The Lower Ventilation Units are powered from the T Buses through the EZC bus and will remaining running through the Phase A and restart when the bus is energized. The Upper Vent Units trip on a Phase A but also will be locked out with load shed. Plausible that the Lower Vent Units also trip since these fans are not routinely used for accident conditions.

D. CORRECT - The Lower Ventilation Units are powered from the T Buses through the EZC bus and will remaining running through the Phase A and restart when the bus is energized. The Upper Vent Units trip on a Phase A but also will be locked out with load shed.

Question ID: RO-C-02800-E9-3 Level/Difficulty: H/3 Comments:

REFERENCE:

01-OHP-4024-103 Drop 2, 3 K/A 022000 K2.01 Containment Cooling System (CCS)

Knowledge of bus power supplies to the following:

Containment cooling fans RO - 3.0 SRO - 3.1 CFR - 41.7 K/A Match Candidate is required to demonstrate knowledge of not only the power supplies for Containment cooling fans but also conditions that isolate the source and actions required to restore them.

Associated objective(s):

(RO-C-02800-E9) Describe the conditions that will cause the following Containment Ventilation component(s) to trip, automatically and/or manually start, and automatically and/or manually reposition.

a. All Fans (022A4.01 - 3.6/3.6)

b. All Containment Isolation Valves (Dampers) (022A4.03 - 3.2/3.2)

c. Air Recirculation/Hydrogen Skimmer Dampers VMO-101/102 (022A4.04 -

3.1/3.2)

d. Containment Ventilation Isolation Signal (022A4.03 - 3.6/4.0)

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 81 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 41 ID: NRCAUDIT07-0737 Points: 1.00 OHP-4021-028-001 Containment Ventilation contains a caution concerning the Loss of Lower Containment Ventilation.

Which ONE of the following describes the concern and the basis for this concern?

The Loss of Lower Containment Ventilation may:

A. cause NESW Containment Isolation Valves to become pressure bound due to temperature changes.

B. cause Phase B isolation signal in about 15 minutes due to Reactor Coolant Pump motors heating the Containment atmosphere.

C. cause the S/G and PZR Level instruments to read low the reference leg temperature will rise.

D. cause Ice Condenser temperatures to rise to unacceptable levels if additional Coolers are not started.

Answer: B Answer Explanation:

A. INCORRECT - This concern is based on isolation of a segment of the NESW system.

B. CORRECT - The RCPs continue to provide heat input to the containment atmosphere during a loss of containment cooling. The heat input would cause a rapid rise in containment pressure, resulting in an SI and CTS actuation based solely on a loss of containment cooling.

C. INCORRECT - The ice condenser door differential pressure caution is contained in the Containment Purge procedure and requires Upper Containment pressure to be less than lower Containment.

D. INCORRECT - Ice condenser cooling is provided by the Glycol cooling system which uses NESW to cool its chillers. A loss of lower containment ventilation should not significantly impact this cooling.

DC COOK OPS Page: 82 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0737 Comments:

REFERENCE:

OHP-4021-028-001 Containment Ventilation Step 3.5, RO-C-02800, Containment Ventilation System K/A 022000 A1.01 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 temperature RO - 3.6 SRO - 3.7 CFR - 41.5 / 45.5 K/A Match Candidate is required to demonstrate not only the ability to monitor containment temperature but also to predict the results from shutting down the CCS.

Associated objective(s):

(RO-C-02800-E11) Describe how Containment Temperature and its associated functional setpoint(s) are used to verify proper operation of the Containment Ventilation system.

{022A4.05 - 3.8/3.8}

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 83 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 42 ID: CM-8332A Points: 1.00 Unit 2 has just reached 100% following a startup from a refueling outage.

Engineering has determined that ONE ICE Condenser INLET Door will NOT OPEN due to a pogo stick left blocking the door.

Which ONE of the following is required to continue operations at power?

REFERENCE PROVIDED A. No action is required since the Ice Condenser Intermediate Deck Doors are all capable of opening.

B. Monitor the ice bed temperature less than or equal to 27oF every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for a maximum of 14 days.

C. Restore the ice bed to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

D. Restore the door to OPERABLE status within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 5 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Answer: D Answer Explanation:

A. INCORRECT - The Inlet doors are still required to be operable.

B. INCORRECT - If the Ice Condenser Intermediate deck doors are inoperable then ITS 3.6.12 Action B applies. This requires temperature to be monitored every 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for up to 14 days.

C. INCORRECT - The ice Condenser temperature is monitored every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per ITS 3.6.11 and 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> are allowed per ITS 3.6.11 if the ice condenser is Inoperable.

D. CORRECT - This is the requirement if an INLET door cannot be fully opened.

DC COOK OPS Page: 84 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: CM-8332A Comments:

REFERENCE:

ITS 3.6.12 pg. 3.6.12-1, 2, & 3.

REFERENCE Provided - Technical Specifications 3.6.11 and 3.6.12 K/A 025000 K6.01 Ice Condenser System Knowledge of the effect of a loss or malfunction of the following will have on the Ice Condenser System:

Upper and lower doors of the ice condenser RO - 3.4 SRO - 3.6 CFR - 41.7 / 45.7 K/A Match Candidate is required to demonstrate knowledge on the effect of Ice Condenser Door unavailability and recognize being Inoperable (1 Hour Tech. Spec.).

Associated objective(s):

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 85 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 43 ID: NRCAUDIT07-0581 Points: 1.00 During a Loss of Coolant Accident (LOCA) there is a time critical action to isolate the Spray Additive Tank Outlet valves in __(1)__ from the time the CTS pumps are restarted after aligning to the recirculation sump. This action is required to __(2)__ .

1 2 A. 17 min stop the injection of nitrogen which would impede depressurization.

B. 5 min stop the injection of sodium hydroxide and prevent raising the recirculation sump pH.

C. 17 min stop the injection of sodium hydroxide and prevent raising the recirculation sump pH.

D. 5 min stop the injection of nitrogen which would impede depressurization.

Answer: B Answer Explanation:

A. INCORRECT - The spray additive tank outlet valve time critical action is 5 min. But the injection of nitrogen is prevented by the spray additive tank outlet valve auto closure prior to lo - lo level.

B. CORRECT - The spray additive tank outlet valve time critical action is 5 min. Sodium hydroxide is added to raise containment sump pH and the isolation time is to prevent exceeding the upper limit.

C. INCORRECT - The spray additive tank outlet valve time critical action is 5 min. Sodium hydroxide is added to raise containment sump pH and the isolation time is to prevent exceeding the upper limit.

D. INCORRECT - The spray additive tank outlet valve time critical action is 5 min. But the injection of nitrogen is prevented by the spray additive tank outlet valve auto closure prior to lo - lo level.

Question ID: NRCAUDIT07-0581 External Topic ID: 0900: Containment Spray/DIS/Hydrogen Recombiner Level/Difficulty: H/3 Comments:

REFERENCE:

RO-C-00900, pg. 21, OHP-4024-105 Drop 3, PMP-4075-TCA-001 DC COOK OPS Page: 86 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 026000 K4.04 Containment Spray System (CSS)

Knowledge of CCS design feature(s) and / or interlock(s) which provide for the following:

Reduction of temperature and pressure in containment after a LOCA by condensing steam, to reduce radiological hazard and protect equipment from corrosion damage (spray)

RO - 3.7 SRO - 4.1 CFR - 41.7 K/A Match Candidate is required to demonstrate knowledge on CSS design features that prevent equipment damage from corrosion during Containment Spray and station critical actions to be taken if design feature fails.

Associated objective(s):

(RO-C-00900-E12) Given the following Containment Spray system and/or hydrogen removal or monitoring system components, describe the conditions that will cause the 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 (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 87 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 44 ID: RO-C-00900-E12-6 Points: 1.00 Following a LOCA inside containment in which pressure reached 3.4 psig, the operator notes the following indications:

• Containment pressure is 2.7 psig and lowering slowly

• Refueling Water Storage Tank (RWST) level is 25 percent and lowering slowly

• Panel 105, Drop 2, SPRAY ADDITIVE TANK LEVEL AT 4000 GAL, alarm has just actuated Assuming NO operator actions, the operator would observe:

• IMO 215/225 CTS Pump Suction Valves __(1)__,

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

• CTS pumps __(3)__.

1 2 3 A. Closed Closed NOT Running B. Open Open Running C. Closed Open NOT Running D. Open Closed Running Answer: B Answer Explanation:

A. INCORRECT - Pump suction and eductor will not close. Pumps should still be running. This distractor is attractive to the student who believes the Containment Spray system will be automatically secured if Containment pressure lowers below the actuation setpoint.

B. CORRECT - Containment Spray will actuate at 2.8 psig. This will start the pumps and open the Spray additive tank valves. The eductor and suction valves are normally open. The Spray additive tank alarm at 4000 gal is an expected alarm after the actuation and shows that NaOH is being injected. If the tank reached the Lo Lo alarm the spray additive tank valves would have isolated (Valves close 589' 9" and the Lo Lo Level alarm annunciates at 589' 8"). The RWST level of <20% is below the switchover setpoint in OHP-4023-ES-1.3, but the actions to close the suction to the RWST must be performed manually.

DC COOK OPS Page: 88 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 C. INCORRECT - Pump suction will not close. Spray Additive tank would still be open. Pumps should still be running. This distractor is attractive to the student who believes the Containment Spray system will be automatically secured and return to the standby configuration if Containment pressure lowers below the actuation setpoint.

D. INCORRECT - Spray Additive tank and eductor would still be open. The Level Lo alarm is an expected alarm. This distractor is attractive to the student who confuses the Spray Additive Tank Level Lo alarm with the Level Lo Lo isolation setpoint.

Question ID: RO-C-00900-E12-6 Comments:

REFERENCE:

1-OHP-4024-105, Annunciator #105 Response: Containment Spray, Drop 2; SOD -00901-001, Containment Spray and Hydrogen Recombiners, RO-C-00900 pg. 21.

K/A 026000 A3.01 Containment Spray System (CSS)

Ability to monitor automatic operation of the CSS, including:

Pump starts and correct MOV positioning.

RO - 4.3 SRO - 4.5 CFR - 41.7 / 45.5 K/A Match Candidate is required to demonstrate the ability to monitor and predict the automatic operation of CSS pumps and MOVs during system operation.

Associated objective(s):

(RO-C-00900-E12) Given the following Containment Spray system and/or hydrogen removal or monitoring system components, describe the conditions that will cause the 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 (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 89 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 45 ID: NRCAUDIT07-1045 Points: 1.00 The following plant conditions exist:

• The Plant is operating at Full Power.

• Control Rods are Full Out in AUTO.

• URV-110 (Condenser Steam Dump Valve) fails partially open and an AEO is immediately dispatched to locally isolate this valve.

Assume no other operator action occurs.

Given these conditions, what indications will you have in the control room that the AEO has successfully isolated the Steam Dump Valve?

Reactor Coolant System Tavg Reactor Power A. Rises Rises B. Lowers Lowers C. Rises Lowers D. Lowers Rises Answer: C Answer Explanation:

A. INCORRECT - If student thinks Tavg must rise as power rises.

B. INCORRECT - If student thinks power must lower as Tavg rises.

C. CORRECT - The failure of the Steam Dump Valve will result in a rise in steam demand which will raise reactor power and result in a drop in Tavg. As the control rods are in the full out position they will not restore RCS temperature. When the operator isolates the Steam Dump Valve, steam demand will drop and reactor power will lower. Since not as much energy is being extracted from the RCS, Tavg will rise. There is no valve indication that lets the control room crew know that this valve is shut and this is an operationally relevant situation. Tavg and power returned to Normal values after Steam Dump valve was closed.

D.INCORRECT - If student chooses the initial response due to the valve failing open and not the end state after it has been isolated.

DC COOK OPS Page: 90 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-1045 External Topic ID: 5200: 5200, Steam Dumps Comments:

REFERENCE:

RO-C-05200, Steam Dump System K/A 039000 K5.08 Main and Reheat Steam System (MRSS)

Knowledge of the operational implications of the following concepts as they apply to the MRSS:

Effect of steam removal on reactivity RO - 3.6 SRO - 3.6 CFR - 41.5 / 45.7 K/A Match Candidate is required to demonstrate an understanding of how steam demand changes effect reactivity and then relate it plant parameters.

Associated objective(s):

(RO-C-05200-E10) State, with respect to Steam Dump System operation, how the following parameters are affected:

a. RCS temperatures

b. Reactivity feedback effects (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 91 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 46 ID: RO-C-05501-E5-3 Points: 1.00 Given the following plant conditions:

• Unit 2 is at 80% power, with all controls systems in normal alignment.

• The Unit 2 Feedwater Pump Discharge Header Pressure Transmitter 2-FPC-250A fails lower to 900 psig during normal plant operation.

(Assume 2-FPC-250A is NOT identified as failed by Digital Control System (DCS).)

Unit 2 Steam Generator levels will:

A. stabilize at lower than normal levels.

B. stabilize at higher than normal levels.

C. initially lower, and then stabilize at normal levels.

D. initially rise, and then stabilize at normal levels.

Answer: D Answer Explanation:

A. INCORRECT - A lower INDICATED FW Discharge Pressure would cause the FW Pump to speed up and raise levels.

B. INCORRECT - The FW Regulating Valves will close down to restore levels to program.

C. INCORRECT - Levels will initially rise before the FW Regulating Valves close down to restore levels to program.

D. CORRECT - When FPC-250A fails downward to 900 psig (Normally ~990), the Feed Pump Speed Control DP initially lowers, causing the control system to raise feedwater pump speed, causing more feed flow. As SG level rises, the Feedwater Regulating Valves close to control level and return level to normal.

Question ID: RO-C-05501-E5-3 External Topic ID: 5501: Main Feed Pump Turbine Controls Level/Difficulty: H/3 Comments:

REFERENCE:

RO-C-05501 K/A 059000 K3.03 Main Feedwater (MFW) System Knowledge of the effect that a loss or malfunction of the MFW System will have on the following:

S/Gs RO - 3.5 SRO - 3.7 CFR - 41.7 / 45.6 DC COOK OPS Page: 92 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A Match Candidate is required to demonstrate knowledge on the effect a malfunction on the MFP Control System will have on the SGs.

Associated objective(s):

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

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 93 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 47 ID: RO27AUDIT-46 Points: 1.00 Given the following conditions on Unit 1:

• A plant startup is in progress.

• The East and West Motor Driven Auxiliary Feedwater Pumps (MDAFPs) are supplying each of their respective steam generators at 80 kpph.

• East and West MDAFP motor amps and discharge pressures are the same.

Comparing the response of the East MDAFP to the West MDAFP, which ONE of the following describes the effects on motor amps and discharge pressure of throttling closed the East MDAFP discharge valves to the SGs?

Motor Amps Pump Discharge Pressure A. East Pump greater than West Pump East Pump greater than West Pump B. East Pump greater than West Pump West Pump greater than East Pump C. West Pump greater than East Pump East Pump greater than West Pump D. West Pump greater than East Pump West Pump greater than East Pump Answer: C Answer Explanation:

A. INCORRECT - Since flow is lower, the East Pump motor amps would be lower.

B. INCORRECT - East pump amps would be lower (less flow) and discharge pressure would be higher (raised system resistance on pump curve).

C. CORRECT - Since flow is lower, the East Pump motor amps would be lower. Discharge pressure would be higher due to raised system resistance on pump curve.

D. INCORRECT - East pump discharge pressure would be higher due to raised system resistance on pump curve.

DC COOK OPS Page: 94 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO27AUDIT-46 Level/Difficulty: H/2 Comments:

REFERENCE:

RO-C-GF17 pg. 62-63, 113-118 K/A 061000 K5.03 Auxiliary / Emergency Feedwater (AFW) System Knowledge of the operational implications of the following concepts as they apply to the AFW System:

Pump head effects when control valve is shut RO - 2.6 SRO - 2.9 CFR - 41.5 / 45.7 K/A Match Candidate is required to demonstrate knowledge of the operational implications of pump head effects when control valve is shut as they apply to the AFW System.

Associated objective(s):

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 95 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 48 ID: NRCAUDIT07-0086 Points: 1.00 Unit 1 is in Mode 3. The 4160 VAC distribution system is being supplied by the Reserve Auxiliary Transformers (RATs). Due to a system disturbance, indicated voltage on the safeguards buses drops.

The following conditions exist:

• T11A Voltage Indication is 112 Volts

• T11B Voltage Indication is 114 Volts

• T11C Voltage Indication is 113 Volts

• T11D Voltage Indication is 114 Volts Which ONE of the following describes the FINAL plant response if voltage remains at these values for an extended period?

__________ will be energized by their respective EDG.

A. All safeguards busses B. T11A and T11C busses C. T11A and T11B busses D. Only T11A bus Answer: C Answer Explanation:

A. INCORRECT - Plausible if the operator mistakenly believes that the 112 volt condition on T11C will cause a degraded bus voltage signal. In fact the signal is generated by 113 volts and lowering on only the A or D 4kV T-Bus.

B. INCORRECT - Plausible if the operator mistakenly believes that degraded bus voltage will load individual busses to the EDG. Degraded bus voltage signals take input from the A and D 4kV T-Busses, and will load one TRAIN of T-Busses (i.e.

A&B or C&D) to the EDG at the setpoint of 113 Volts and lowering.

C. CORRECT - An Under voltage condition of 113 V will energize 62-1 T11A. After a 111 Second delay it will open T11A9 and T11B1 causing T11 A and T11B to lose power. This will cause the EDG to start and energize T11A and T11B.

D. INCORRECT - Plausible if it is thought that only A or D T-Busses will energize from the EDG, since they are the only busses that detect voltage for the degraded bus voltage signal.

DC COOK OPS Page: 96 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0086 External Topic ID: 8201: ESS Electrical Distribution Comments:

Reference:

RO-C-08201, Engineered Safety Systems Electrical pg. 25-26, and Handout 3

Annunciator #121 Response, Drop 78 Train B Aux Buses Under voltage K/A 062000 K1.02 A.C. Electrical Distribution System Knowledge of the physical connections and/or cause-effect relationships between the A.C.

Distribution System and the following systems:

ED/G RO - 4.1 SRO - 4.4 CFR 41.2 to 41.9 / 45.7 / 45.8 K/A Match Candidate is required to determine impact on DG and AC Bus based on degraded bus voltages.

Associated objective(s):

(RO-C-08201-E6) Describe all inputs, indications, and outputs for the following circuit/controls:

a. Load Shed

b. Load Conservation

c. Degraded Bus Protection

d. Synchronization Instrument and Controls (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 97 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 49 ID: RO-C-08201-E4-1 Points: 1.00 Unit 2 was operating at about 100% power when a Complete Loss of Onsite and Offsite AC power occurred 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ago. Unit 2 dispatched operators after 30 minutes 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.

• SG pressures have been stabilized.

• While performing Step 36, power could NOT be restored to the battery charger for the 2CD 250VDC buses.

• The actions of step 36 to restore chargers for Train N and AB, 600V AC Busses and Control Room Cooling were successfully performed.

When the Train CD 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 Train A ECCS pumps from the control room will be lost.

C. the ability to start and stop Train B ECCS pumps from the control room will be lost.

D. the ability to control AFW flow to SG 2 & 3 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 and are powered from Train N B. CORRECT - A loss of DC control power will prevent breaker operations with the control switch (and trip functions).

C. INCORRECT - A loss of DC control power will prevent breaker operations with the control switch (and trip functions) but CD is on Train A.

D. INCORRECT - Plausible since many AOVs will fail closed on loss of DC and the TDAFW Pump Valves use DC but will fail as is and ALL 4 are powered from Train N.

The East AFW pump which feeds SG 2 & 3 is feed from the Train A Bus and will be impacted.

DC COOK OPS Page: 98 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-08201-E4-1 External Topic ID: 8201: ESS Electrical Distribution Comments:

REFERENCE:

RO-C-08201, Engineered Safety System Electrical Distribution System K/A 063000 K2.01 D.C. Electrical Distribution System Knowledge of bus power supplies to the following:

Major DC loads RO - 2.9 SRO - 3.1 CFR - 41.7 K/A Match Candidate is required to determine the impact the loss of DC has on major loads.

Associated objective(s):

(RO-C-08201-E4) Describe the purpose of the systems/components that interface with and/or provide a support function for the ESS Electrical Distribution System and explain how a loss of the support system will affect its operation.

a. DC Power

b. 4kV Power Supplies (RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 99 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 50 ID: RO-C-0820360101-E1-1 Points: 1.00 The following plant conditions exist:

• Unit 1 is in Hot Standby.

• Reserve Feed Breaker 12AB has tripped due to a fault.

• 1AB Emergency Diesel Generator failed to start.

• Ann.119, Drop 9, BATTERY CHARGER 1AB1 FAILURE, is LIT.

Which ONE of the following describes the response of the ESFAS controlled and actuated systems resulting from these conditions?

The Train B Solid State Protection output relays will:

A. function normally. The DC powered valves will reposition but the Train B AC powered valves and pumps will NOT reposition or start as desired.

B. NOT function normally. ALL of the automatic actions for the Train B equipment will have to be manually performed.

C. function normally. The Train B DC and AC valves will reposition as desired. The Train B pumps will NOT restart as required.

D. NOT function normally. NO Train B equipment will be aligned by the crew since power is not available.

Answer: A Answer Explanation:

A. CORRECT - The SSPS output relays will be powered from the CRIDs since they are still aligned to DC Battery AB. The DC valves will reposition as desired but the AC motors and pumps will not operate.

B. INCORRECT - The SSPS output relays will be powered from the CRIDs since they are still aligned to DC Battery AB. The actions to close the DC valves would be performed if the SSPS relays did not function. The AC valves and pumps still have no power.

C. INCORRECT - The SSPS output relays will be powered from the CRIDs since they are still aligned to DC Battery AB. The AC valves and pumps still have no power.

D. INCORRECT - The SSPS output relays will be powered from the CRIDs since they are still aligned to DC Battery AB. The actions to close the DC valves would be performed if the SSPS relays did not function.

DC COOK OPS Page: 100 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-0820360101-E1-1 Level/Difficulty: H/3 Comments:

REFERENCE:

SD-08204 250 VDC Distribution, RO-C-08204 250 VDC Distribution K/A 064000 K3.02 Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following:

ESFAS controlled or actuated systems RO - 4.2 SRO - 4.4 CFR: 41.7 / 45.6 K/A Match Candidate is required to demonstrate knowledge of how ESFAS systems are impacted by loss of EDG.

Associated objective(s):

(RO-C-0820360101-E1) Identify the event and predict the response of the plant to Loss of CRID Instrument Inverters and Power Panels, including final plant configuration.

(RO31 NRC Exam 2018 26-50)

DC COOK OPS Page: 101 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 51 ID: RO-S-EOP0310512-T4-1 Points: 1.00 Given the following conditions:

Unit 2 experienced a generator trip followed by a failure of Bus 2D to transfer to reserve power.

2CD Emergency Diesel Generator (EDG) is running supplying power to T21C and T21D.

• Reserve Power has been restored to Bus 2D

• 2-OHP-4023-SUP-002 Restoration of Reserve Power to 4KV Buses, Attachment J Energizing Bus T21D from Bus 2D is in progress The crew is preparing to parallel DG2CD which is carrying bus T21D to Bus 2D.

Which ONE of the following conditions is verified prior to closing the T21D12 Bus 2D supply to bus T21D?

A. Synchroscope is moving SLOWLY in the SLOW direction.

Adjust Bus 2D (Run) voltage 2-3 volts higher than DG2CD (Start) voltage.

B. Synchroscope is moving SLOWLY in the SLOW direction.

Adjust DG2CD (Start) voltage 2-3 volts higher than Bus 2D (Run) voltage.

C. Synchroscope is moving SLOWLY in the FAST direction.

Adjust Bus 2D (Run) voltage 2-3 volts higher than DG2CD (Start) voltage.

D. Synchroscope is moving SLOWLY in the FAST direction.

Adjust DG2CD (Start) voltage 2-3 volts higher than Bus 2D (Run) voltage.

Answer: D Answer Explanation:

A. INCORRECT - This would cause the DG to slightly reverse power and try to take negative VARS. Plausible since this is the opposite direction of the normal parallel when a DG is paralleled to the bus.

B. INCORRECT - This would cause the DG to slightly reverse power while still maintaining positive VARS. Plausible if candidate believes the bus must pick up load from the DG.

C. INCORRECT - This would cause a slight negative VARS. Typically the source coming onto the bus is at a higher voltage. Synchroscope direction is correct.

DC COOK OPS Page: 102 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. CORRECT - The procedure requires the Operator to ensure that the Synchroscope 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. DG voltage is slightly higher to maintain positive VARs Question ID: RO-S-EOP0310512-T4-1 Level/Difficulty: H/2 Comments:

REFERENCE:

2-OHP-4023-SUP-002 Restoration of Reserve Power to 4KV Buses, Attachment J Energizing Bus T21D from Bus 2D K/A 064000 A4.09 Emergency Diesel Generator (ED/G) System Ability to manually operate and/or monitor in the control room:

Establishing power from the ring bus (to relieve ED/G)

RO - 3.2 SRO - 3.3 CFR - 41.7 / 45.5 to 45.8 K/A Match Candidate is required to determine the correct actions for restoring AC power to the bus to relieve the EDG.

Associated objective(s):

(RO-C-EOP14-E3) Discuss the plant response to a loss of all AC power event with and without operator action.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 103 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 52 ID: CM-7832 Points: 1.00 Which ONE of the following actions will occur due to a HIGH radiation alarm on monitor RRS-1001, Liquid Radwaste Effluent Line?

A. Alarms at WDS Panel and Control Rooms ONLY- NO automatic actions.

B. 12-RRV-285, Liquid Waste Disposal Effluent Discharge Header Shutoff Valve closes.

C. 1-RRV-286 or 2-RRV-287, Liquid Waste Effluent To Header CIRC Water Discharge Header Shut Off Valve closes.

D. 12-RRV-284, Liquid Release Control Valve closes.

Answer: B Answer Explanation:

A. INCORRECT - The HIGH rad. level causes the isolation. The ALERT level just gives an alarm.

B. CORRECT - The HIGH alarm on RRS-1001 will cause RRV-285 to close and also trip the WECT and monitor tank pumps.

C. INCORRECT - RRV-286 and 287 are used to select which unit to discharge to and will close on loss of CW flow. They do not auto close on High radiation.

D. INCORRECT - RRV-284 controls the release flow rate but does not isolate on HIGH radiation.

Question ID: CM-7832 External Topic ID: 2200: 2200, Liquid Rad Waste1350: 1350, Radiation Monitoring System Level/Difficulty: F/3 Comments:

REFERENCE:

12-OHP-4024-139, Drop 18 K/A 073000 K4.01 Process Radiation Monitoring (PRM) System Knowledge of PRM System design feature(s) and/or interlock(s) which provide for the following:

Release termination when radiation exceeds setpoint RO - 4.0 SRO - 4.3 CFR - 41.7 DC COOK OPS Page: 104 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A Match Candidate is required to demonstrate knowledge on the interlock design for high radiation release termination.

Associated objective(s):

(RO-C-02200-E8) Describe the conditions that will cause the following liquid waste disposal component to trip, automatically/manually start and/or automatically/manually reposition.

a. RRV-285 (RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 105 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 53 ID: RO-C-01900-E9-1 Points: 1.00 Which ONE of the following conditions describes to operation of the Normal and Alternate ESW supply valves to the Diesel Generators?

A. The Normal ESW supply valves are normally open and receive an open signal on a DG start. The Alternate ESW Valves are normally closed but receive an open signal on low ESW flow to the DG.

B. The Normal ESW supply valves are normally closed and receive an open signal on a DG start. The Alternate ESW Valves are normally closed and must be manually opened on low ESW flow to the DG.

C. The Normal ESW supply valves are normally open. They do NOT have any automatic features. The Alternate ESW Valves are normally closed but receive and open signal on low ESW flow to the DG.

D. The Normal ESW supply valves are normally open and receive an open signal on a DG start. The Alternate ESW Valves are normally closed and must be manually opened on low ESW flow to the DG.

Answer: D Answer Explanation:

A. INCORRECT - The Normal ESW supply valves are normally open and receive and open signal. The alternate valves do NOT receive any open signals.

B. INCORRECT - The Normal ESW supply valves are normally open and receive and open signal. The alternate valves do NOT receive any open signals and must be manually opened if required.

C. INCORRECT - The Normal ESW supply valves are normally open and receive and open signal. The alternate valves do NOT receive any open signals.

D. CORRECT - The Normal ESW supply valves are normally open and receive and open signal. The alternate valves do NOT receive any open signals.

DC COOK OPS Page: 106 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-01900-E9-1 Level/Difficulty: H/2 Comments:

REFERENCE:

OP-1-98416 K/A 076000 K1.21 Service Water System (SWS)

Knowledge of the physical connections and/or cause-effect relationships between the SWS and the following systems:

Auxiliary backup SWS RO - 2.7 SRO - 2.9 CFR - 41.2 to 41.9 / 45.7 to 45.8 K/A Match Candidate is required to demonstrate knowledge of the operation of the Alternate ESW supply to the EDG.

Associated objective(s):

(RO-C-01900-E9) Given the following Essential Service Water System components, 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 (RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 107 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 54 ID: RO-C-EOP08-E18-3 Points: 1.00 Unit 1 has experienced a Reactor Trip. A Safety Injection was actuated when a SG Tube rupture was identified. The crew has performed the steps of E-0 and has transitioned to 1-OHP-4023-E-3 Steam generator Tube Rupture. The Crew has reset SI and Phase A Containment Isolation and has begun to restore Control Air to Containment by placing control switches for XCR-100, 101, 102, and 103 to the open position.

Which ONE of the following describes the reason for checking the Containment CTRL Air Header Pressure Low Annunciator on Panel 204 Drop 33 and the subsequent actions required?

If the alarm is LIT:

A. air pressure will NOT be available for the RCS depressurization. A transition to 1-OHP-4023-ECA-3.3, SGTR Without Pressurizer Pressure Control will be required.

B. an air leak inside containment could impact the containment accident analysis. The XCR valves must be reclosed.

C. the XCR valves have failed to fully open. An operator must be dispatched to locally open the containment XCR isolation valves.

D. air pressure will NOT be available to establish letdown. 1-OHP-4023-ES-3.1, Post- SGTR Cooldown using Backfill is NOT an option, so secondary system contamination minimization steps will be required.

Answer: B Answer Explanation:

A. INCORRECT - The Pressurizer PORVs, NRV-152 and NRV-153, have backup Air Supplies.

B. INCORRECT - The alarm check is not used to verify opening of the XCR valves. An operator would not be dispatched to open the valves inside containment.

C. INCORRECT - Letdown may not be available if air could not be restored, but this is not the reason for this check. The steps to minimize secondary radiation are performed for all Cooldown methods.

D. CORRECT - An Air leak inside the ice condenser containment could raise the amount of gasses within containment and challenge the accident analysis. The air supply headers are isolated if the alarm is lit indicating a potential air leak.

DC COOK OPS Page: 108 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-EOP08-E18-3 Level/Difficulty: F/3 Comments:

REFERENCE:

12-OHP-4023-E-3 Step 12 and background.

K/A 078000 2.4.31 Instrument Air System (IAS)

Emergency Procedures/Plan Knowledge of annunciator alarms, indications, or response procedures.

RO - 4.2 SRO - 4.1 CFR - 41.10 / 45.3 K/A Match Candidate is required to demonstrate knowledge of the reason for the annunciator check in the emergency procedures for the Containment air system.

Associated objective(s):

(RO-C-EOP08-E18) For the E-3 series procedures and the ECA-3 series procedures discuss the basis or reason for all Steps.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 109 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 55 ID: NRCAUDIT07-1073 Points: 1.00 A malfunction of the Containment Chilled Water (CHW) system has resulted in a partial loss of 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. Pressure is 0.18 PSIG, Upper containment temperature 97°F, Lower containment temperature 110°F.

B. Pressure is 0.35 PSIG, Upper containment temperature 105°F, Lower containment temperature 97°F.

C. Pressure is 0.15 PSIG, Upper containment temperature 105°F, Lower containment temperature 105°F.

D. Pressure is -0.05 PSIG, Upper containment temperature 115°F, Lower containment temperature 115°F.

Answer: A Answer Explanation:

A. CORRECT - 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.

B. INCORRECT - Pressure too high & Upper temp too high, plausible since upper temp is below lower temp limit & Pressure is still reasonable.

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

D. INCORRECT - Upper Temp is too high.

DC COOK OPS Page: 110 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-1073 External Topic ID: 3400: 3400, Containment Level/Difficulty: F/3 Comments:

REFERENCE:

Unit 2 T.S. 3.6.4 and 3.6.5 K/A 103000 A1.01 Containment System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the Containment System controls including:

Containment pressure, temperature, and humidity RO - 3.7 SRO - 4.1 CFR - 41.5 / 45.5 K/A Match Candidate is required to demonstrate knowledge of the Containment limits that are being monitored.

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.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 111 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 56 ID: RO-C-01200-E9-1 Points: 1.00 Which ONE of the following describes the operation of the ROD Step Counters.?

Following a reactor trip the Step counters will:

A. reset to zero when the trip breakers open.

B. indicate the position indicated just before the trip. Closing the trip breakers will reset the Step Counters.

C. indicate the position indicated just before the trip. The Step Counters are reset from the Control Room just prior to a reactor Startup.

D. indicate the position indicated just before the trip. The Step Counters are reset in the rod drive cabinets just prior to closing the reactor trip breakers.

Answer: C Answer Explanation:

A. INCORRECT - The Step Counters indicate the Demand Position of the Rods.

B. INCORRECT - The Step Counters indicate the Demand Position of the Rods. They will NOT reset when the breakers are closed.

C. CORRECT - The Step Counters indicate the Demand Position of the Rods prior to the trip.

The Step Counters are reset from the Full Length Step Counter Reset Switch on the Control Board.

D. INCORRECT - The Step Counters indicate the Demand Position of the Rods prior to the trip.

The Step Counters are reset from the Full Length Step Counter Reset Switch on the Control Board.

Question ID: RO-C-01200-E9-1 Level/Difficulty: F/2 Comments:

REFERENCE:

1-OHP-4021-001-002 Reactor Start-Up K/A 001000 K4.12 Control Rod Drive System Knowledge of CRDS design feature(s) and/or interlock(s) which provide for the following:

Re-zeroing rod demand position counters RO - 2.5 SRO - 2.6 CFR - 41.7 DC COOK OPS Page: 112 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/a Match Candidate is required to demonstrate knowledge of the step counter operation on a reactor trip and how they are re-zeroed Associated objective(s):

(RO-C-01200-E9) DESCRIBE how operation of the full-length step counter reset switch affects the Rod Control System.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 113 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 57 ID: CM-6351 Points: 1.00 Unit 1 is at 100% power with all control systems in automatic. Power Range channel N42 upper detector output rapidly fails low.

Which ONE of the following describes the expected plant response with no operator actions?

A. The Channel 2 OT Delta-T Setpoint will be LOWER than it was before the failure.

B. Rods will WITHDRAW until the Tavg - Tref error offsets the Power Mismatch circuit.

C. The Channel 2 OP Delta-T Setpoint will be HIGHER than it was before the failure.

D. Rods will continuously INSERT until the operator places Rod Control in MANUAL.

Answer: A Answer Explanation:

A. CORRECT - The Delta flux penalty in the OTDT setpoint calculation subtracts - fDI.

When qt-qb is more negative than -37% (as it would be for 0-50) then fDI =

-.33(37% + -50%) or -.33(-13%)= 4.29. This is the value which is subtracted to make the setpoint lower, as it should be if the axial power skew were this pronounced. (See COLR Figure 5)

B. INCORRECT - Plausible if student believes that power rate mismatch circuit would act to increase reactor power since the upper detector failed low. However since this channel would not be the auctioneered high NI channel, it is not input to the rod control circuit.

C. INCORRECT - OPDT setpoint calculation zeroes out the Delta flux penalty. Plausible if student believes that the OPDT setpoint will be higher since power is lower.

D. INCORRECT - Plausible if student believes that an upper detector channel failure will result in a rod insertion command. If the upper detector had failed high, then this statement would be true.

Question ID: CM-6351 External Topic ID: 1100-1: 1100-1, RPS/ESFAS Signals Comments:

REFERENCE:

T.S. 3.3.1 RTS Instrumentation DC COOK OPS Page: 114 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 015000 K6.01 Nuclear Instrumentation system Knowledge of the effect of a loss or malfunction of the following will have on the NIS:

Sensors, detectors, and indicators RO - 2.9 SRO - 3.2 CFR - 41.7 / 45.7 K/A Match Candidate is required to demonstrate knowledge of how the detector failure will impact the AFD input and the OTDT circuit.

Associated objective(s):

(RO-C-AOP0300412-E1) identify the event and predict the response of the plant to a Power Range Nuclear Instrument Malfunction, including final plant configuration.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 115 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 58 ID: RQ-C-05101-T4-3A Points: 1.00 Given the following plant conditions on Unit 1:

• The unit is at 100% power and stable.

• Steam Generator Level Controls are in AUTOMATIC.

• Steam Generator #12 Feed Flow Channel 1, 1-FFC-220, is selected to the Steam Generator Level Control System.

If 1-FFC-220 instantaneously fails off-scale low, which ONE of the following describes the expected plant response?

Note: Assume no operator action until the procedure entry.

Procedures:

1-OHP-4022-013-015 Feedwater Flow Instrument Malfunction 1-OHP-4022-IFR-001 Instrument Failure Response The Steam Generator Level Control system will:

A. initially raise feed flow and then slowly return SG#12 level to approximately program level.

Enter 1-OHP-4022-013-015 Directly to address the failed channel.

B. automatically transfer the SG#12 FW Regulating Valve Controller to Manual to maintain the current valve position. Verify the plant is stable per 1-OHP-4022-IFR-001 and then enter 1-OHP-4022-013-015 to address the failed channel.

C. initially lower feed flow and then slowly return SG#12 level to approximately program level.

Verify the plant is stable per 1-OHP-4022-IFR-001 and then enter 1-OHP-4022-013-015 to address the failed channel.

D. automatically transfer the SG#12 FW Regulating Valve Controller to Feed Flow Channel 2, 1-FFC-221 to allow continued automatic operation. Enter 1-OHP-4022-013-015 Directly to address the failed channel.

Answer: B Answer Explanation:

A. INCORRECT - The FRV Controller will shift to manual. This is the response to a slow Steam Flow failure. Plausible if student believes the controller continues to respond to the failed input.

B. CORRECT - A sudden failure off scale will cause the Taylor Controller to shift to manual at the current position. (The SGWLC will fail-over to the value 2 seconds prior to the incident)

C. INCORRECT - The FRV Controller will shift to manual. This is the response to a slow FW Flow failure. Plausible if student believes the controller remains in automatic and misunderstands the SF-FF controller action.

DC COOK OPS Page: 116 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. INCORRECT - The FRV Controller will shift to manual. This is the response to failures of instrumentation associated with the Digital Control Systems. Plausible if student believes the system automatically selects a new good input.

Question ID: RQ-C-05101-T4-3A External Topic ID: 5100-2: SGWLC Comments:

REFERENCE:

RO-C-05100, Steam Generator System TP 108 K/A 016000 A2.03 Non-Nuclear Instrumentation System (NNIS)

Ability to (a) predict the impacts of the following malfunctions or operations on the NNIS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations.

RO - 3.0 SRO - 3.3 CFR - 41.5 / 43.5 / 45.35 / 45.

K/A Match Candidate is required to demonstrate the ability to predict impact of loss of transmitted signal and identify the procedure to mitigate the consequences of failure.

Associated objective(s):

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 117 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 59 ID: RO-C-ADM04-E5-03 Points: 1.00 Given the following data obtained from a scheduled surveillance.

From Technical Data Book figure 19.1. - All times in seconds:

VALVE DIRECTION MIN MAX LIMIT REF STROKE STROKE 1-VCR-103 CLOSED 1.6 4.8 5.0 3.2 1-VCR-104 CLOSED 1.6 4.6 5.0 3.1 Stroke Time Test Results:

1-VCR-103 CLOSED: 4.7 seconds 1-VCR-104 CLOSED: 4.8 seconds Which ONE of the following describes the status of these Containment Isolation Valves?

A. Both valves are OPERABLE; ALL stroke times are within surveillance requirements.

B. 1-VCR-104 is INOPERABLE, immediate retesting is not allowed for surveillances.

C. Both valves are OPERABLE, 1-VCR-104 requires successful immediate retest required for continued OPERABILITY.

D. Both valves are INOPERABLE, until a retest yields pass results.

Answer: C Answer Explanation:

A. INCORRECT - Plausible if the student uses Limit time of 5 seconds with the as the only requirement. INCORRECT because the valve is outside MAX time for CLOSED, but under the LIMIT per OHI-4016. Must be declared INOPERABLE if cannot retest immediately.

B. INCORRECT - Plausible because OHI-4016 does NOT allow retesting valve for valves exceeding the LIMIT stroke time. INCORRECT because OHI-4016 does allow for retesting valves that exceed the MAX stroke time.

C. CORRECT - 1-VCR-104 Valve is outside the MAX time for CLOSED, but under the limit per OHI-4016. Must be declared INOPERABLE if cannot retest immediately.

D. INCORRECT - Plausible if the student uses the max stroke or reference times to determine operability. INCORRECT because VCR-103 would remain operable even if a retest were not performed. VCR -104 can remain OPERABLE if it can be retested and pass the max stroke on the second timing.

DC COOK OPS Page: 118 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-ADM04-E5-03 Level/Difficulty: H/3 Comments:

REFERENCE:

OHI-4016 Conduct of Operations Guidelines K/A 029000 2.2.12 Containment Purge System (CPS)

Equipment Control Knowledge of surveillance procedures.

RO - 3.7 SRO - 4.1 CFR - 41.10 / 45.13 K/A Match Candidate is required to demonstrate knowledge of surveillance requirements for Containment Purge valve stroke times.

Associated objective(s):

(RO-C-ADM04-E5) Describe the theory and practice of valve testing.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 119 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 60 ID: CM-0385 Points: 1.00 Given the following plant conditions:

• Unit 2 is at 100% power.

• All systems are aligned in Normal, controls in Automatic.

• A leak develops in the reference leg for the controlling Narrow Range SG Level channel (2-BLP-111)

• The condensing pot cannot keep up with the leak.

Assume NO operator action is taken.

Which ONE of the following best describes the response of the plant?

A. The turbine will trip on Hi-Hi SG Level.

B. The reactor will trip on Low-Low SG Level.

C. Water Level in #1 SG will begin to rise and then return to Program level as a Level Error develops.

D. Water level in #1 SG will begin lowering and then return to program when a Steam Flow/Feed Flow error develops and equals the Level Error.

Answer: B Answer Explanation:

A. INCORRECT - Decreasing mass in the reference leg will cause Indicated level to rise. Actual level, however, will be decreasing due to the FRV responding to the erroneous channels indication.

B. CORRECT - Reference leg leak results in rising indicated level, shutting of FRV to control level, and the subsequent lowering of Actual level to the trip setpoint.

C. INCORRECT - Water level will rise, but the level error (from the failed channel) is constantly indicating High level.

D. INCORRECT - Water level does actually lower, but the Steam Flow/Feed Flow error is NOT integrated to a value that will counteract the Level Error.

Question ID: CM-0385 Comments:

REFERENCE:

GF27 Sensors and Detectors DC COOK OPS Page: 120 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 035000 A3.01 Steam Generator System (S/GS)

Ability to monitor automatic operation of the S/G, including:

S/G water level control RO - 4.0 SRO - 3.9 CFR - 41.7 / 45.5 K/A Match Candidate is required to demonstrate ability to monitor and predict the response of SG water level control to a reference leg break.

Associated objective(s):

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 121 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 61 ID: NRCAUDIT07-1032 Points: 1.00 The following plant conditions exist on Unit 2:

• The reactor is critical at 11% power just prior to rolling the turbine.

• Manual FW control is maintaining SG Levels.

• RCS temperature is being controlled by steam dumps in automatic in the Steam Pressure Mode.

UPC-101 Steam Header Pressure transmitter fails to 0 psig.

With NO operator action, which ONE of the following statements describes the resulting SG Level response?

Steam Generator levels will initially:

A. lower due to shrink and then rise as RCS temperature raises.

B. rise due to swell and then lower due to the higher steam release rate.

C. lower due to shrink and then rise due to higher FW flow at the lower pressure.

D. rise due to swell and then lower due to lower FW flow at the high pressure.

Answer: A Answer Explanation:

A. CORRECT - The Steam Dumps will be trying to maintain Steam Line pressure. When the input pressure fails low the dumps will close, causing SG pressure to be higher. This will cause an initial shrink followed by level rising as the RCS temperature raises.

B. INCORRECT - The failure will cause steam dumps to close.

C. INCORRECT - The SG will shrink but due to higher pressure.

D. INCORRECT - The SG will shrink from the higher pressure.

DC COOK OPS Page: 122 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-1032 External Topic ID: 5200: 5200, Steam Dumps Comments:

REFERENCE:

RO-C-05200 Steam Dump System K/A 041000 A4.05 Steam Dump System (SDS) and Turbine Bypass Control Ability to manually operate and / or monitor in the control room:

Main steam header pressure K/A Match Candidate is required to demonstrate ability to monitor the actual steam header pressure from the effects on SG and RCS during an instrument failure.

Associated objective(s):

(RO-C-05200-E7) Given a set of initial plant conditions and any failure of the instrumentation or controls listed below, state the effect upon and final conditions of the Steam Dump System.

a. Condenser Vacuum and Circulating Water Pump breaker position

b. Controllers and positioners

c. MPC-253

d. MPC-254

e. Tavg channels

f. UPC-101 (RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 123 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 62 ID: RO-C-01100-E2-3 Points: 1.00 Which one of the following describes purpose and reason for the Turbine Trip/Reactor Trip?

The Turbine Trip/Reactor Trip protects against:

A. the loss of secondary heat removal. The trip is NOT required below 30% since the plant is designed for a 40% loss of load using steam dumps.

B. a RCS overpressure condition. The trip is NOT required below 30% since the plant is designed for a 40% loss of load using control rods only.

C. the loss of secondary heat removal. The trip is NOT required below 30% since OT Delta T trip provides a variable setpoint based on Tave and RCS Pressure.

D. excessive RCS Cooldown. The trip is NOT required below 30% since the RCS Delta T is small at these power levels.

Answer: A Answer Explanation:

A. CORRECT - Plant can handle full loss of load. RCS high pressure is backup. Trip is in anticipation of loss of secondary heat sink.

B. INCORRECT - Plant can handle full loss of load. RCS high pressure is backup. Plant can handle full load rejection. Rods are typically credited with 10% and dumps with 40%.

C. INCORRECT - Correct purpose but he OTDT would only provide protections at higher powers.

D. INCORRECT - Correct reason the turbine is tripped from the reactor trip. - This is not blocked above 30%. The DT would be lower but not a correct reason.

Question ID: RO-C-01100-E2-3 Level/Difficulty: F/4 Comments:

REFERENCE:

T.S. Bases 3.3.1 Function 16 Turbine Trip DC COOK OPS Page: 124 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 045000 K5.18 Main Turbine Generator (MT/G) System Knowledge of the operational implications of the following concepts as they apply to the MT/G System:

Purpose of low-power reactor trips (limited to 25% power)

RO - 2.7 SRO - 3.2 CFR - 41.5 / 45.7 K/A Match Candidate is required to demonstrate knowledge of the purpose of turbine generator / reactor trip that is blocked at low powers and the plant impact.

Associated objective(s):

(RO-C-01100-E2) Explain the purposes/functions/bases of the following RPS actuation circuits (Reactor Trips):

a. Manual

b. Power Range (high, low, and rate)

c. Intermediate Range

d. Source Range

e. OT(T

f. OP(T

g. Pressurizer Pressure (high and low)

h. Pressurizer Level

i. Loss of RCS Flow

j. SG Low-Low Level

k. SG Low Level w/ Feed/Steam Flow Mismatch

l. RCP UV

m. RCP UF

n. RCP Breaker Position

o. Turbine Trip (oil pressure and stop valve position)

p. Safety Injection (RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 125 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 63 ID: RO-C-02300-E8-2 Points: 1.00 During a release of the #1 Gas Decay tank, the 12-RRV-305, GDT Release Header to Aux Bldg.

Vent Stack Pressure Reducing Valve fails partially open raising the regulated pressure from 5 psig to 20 psig.

Which ONE of the following describes the impact of this failure on the release of #1 Gas Decay Tank and the required actions, if any?

A. The release flowrate will be 4 times higher. This higher flowrate will cause the VRS-1505 Unit 1 Vent Effluent Monitor to incorrectly read a higher activity level. Another Unit 1 Vent fan should be started.

B. The release flowrate will be 4 times higher. This higher flowrate will NOT change the reading on VRS-1505 Unit 1 Vent Effluent Monitor since it is sampling the flow path.

C. The release flowrate will be 2 times higher. This higher flowrate will cause the VRS-1505 Unit 1 Vent Effluent Monitor to correctly read a higher activity level. The release should be terminated by closing 12-RRV-306, GDT Release Header To Aux Bldg. Vent Stack Shutoff Valve.

D. Initially the release flowrate will be 2 times higher. 12-RRV-306, GDT Release Header To Aux Bldg. Vent Stack Shutoff Valve controller automatically adjusts to lower the flow to the desired value so it will NOT change the reading on VRS-1505 Unit 1 Vent Effluent Monitor will not be impacted.

Answer: C Answer Explanation:

A. INCORRECT - The pressure is 4 times higher but the flow should change by the square root of the DP. Flow would only be twice as high. VRS-1505 will CORRECTLY read a higher activity since there is less dilution of the release effluent.

B. INCORRECT - The pressure is 4 times higher but the flow should change by the square root of the DP. Flow would only be twice as high. The flow path being sampled will show a higher activity since there is less diluting form the Aux Bldg. vent.

C. CORRECT - Flow should change by the square root of DP. VRS -1505 should show a higher reading and may automatically isolate the release. The release should be terminated.

D. INCORRECT - The RRV-306 is set to a position to give the desired flow. It will not automatically throttle.

DC COOK OPS Page: 126 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-02300-E8-2 External Topic ID: 1350: Radiation Monitoring System Level/Difficulty: H/3 Comments:

REFERENCE:

GF27: Sensors and Detectors K/A 071000 K3.05 Waste Gas Disposal System (WGDS)

Knowledge of the effect that a loss or malfunction of the Waste Gas Disposal System will have on the following:

ARM and PRM systems RO - 3.2 SRO - 3.2 CFR - 41.7 / 45.6 K/A Match Candidate is required to demonstrate knowledge of how a malfunction in the Waste Gas release flowpath will impact the monitors.

Associated objective(s):

(RO-C-02300-E3) Describe the function of the following Gaseous Waste Disposal System Components:

a. DCR-201, 203

b. DCR-202, 204

c. DCR-207

d. N-160

e. RRV-304

f. RRV-305

g. RRV-306

h. Automatic Gas Analyzer (AGA)

i. Alternate Oxygen Monitor (AOM)

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 127 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 64 ID: CM-3047 Points: 1.00 While performing a liquid release through Unit 2, all Circulating Water Pumps trip.

Which ONE of the following will occur FIRST?

A. The liquid effluent unit isolation valve, 2-RRV-286, shuts.

B. The monitor tank pump trips.

C. The Data Acquisition Module alarms due to high flow.

D. The liquid effluent header isolation valve, RRV-285, shuts.

Answer: A Answer Explanation:

A. CORRECT - The liquid effluent release unit release valve is interlocked with that unit's CW pumps. Unit CW pump breaker must be closed for this valve to be opened.

Trip of all unit CW pumps will close the unit release valve.

B. INCORRECT - The monitor tank pump does not trip when unit CW pumps trip. Plausible if student believes that CW pump interlock includes trip of monitor tank pump.

The monitor tank pump will trip on sample flow out of range. Low sample flow in this case will not result until RRV-286 closes.

C. INCORRECT - The sample flow would not be affected until the unit effluent discharge valve closed, and then low sample flow would alarm. Plausible if student discerns INCORRECT chronology for flow in sample line.

D. INCORRECT - The liquid effluent header isolation valve will trip closed when sample flow is outside of range. This will not happen until the unit isolation valve closes.

Plausible if student discerns INCORRECT chronology for flow in sample line.

Question ID: CM-3047 Comments:

REFERENCE:

SD-02200, Liquid Waste Disposal System DC COOK OPS Page: 128 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 075000 K1.02 Circulating water System Knowledge of the physical connections and / or cause - effect relationships between the Circulating Water System and the following systems:

Liquid Radwaste system K/A Match Candidate is required to demonstrate knowledge of the effects of a CW pump trip on the Liquid release.

Associated objective(s):

(RO-C-02200-E5) Explain how a loss of each of the following support systems will affect the operation of the Liquid Waste Disposal System:

a. Circ Water System

b. Radiation Monitoring System (RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 129 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 65 ID: CM-0628A Points: 1.00 A rupture in the fire protection water main results in pressure dropping to 135 psig for 20 seconds before recovering to 155 psig.

Which ONE of the following describes the automatic fire protection system response that resulted in this recovery?

A. Motor driven fire pump and 'W' Diesel Fire Pump started, 'E' Diesel fire pump remained in standby.

B. Motor Driven Fire Pump and `E' Diesel Fire Pump started, `W' Diesel Fire Pump remained in standby.

C. Motor Driven Fire Pump started, both Diesel Fire Pumps remained in standby.

D. Motor Driven Fire Pump and both Diesel Fire Pumps have started.

Answer: C Answer Explanation:

A. INCORRECT - Electric pump starts if fire protection header is < 140 psig for 2 seconds, West diesel pump starts if fire protection header is < 120 psig for 25 seconds.

B. INCORRECT - Electric pump starts if fire protection header is < 140 psig for 2 seconds, east diesel pump starts if fire protection header is < 130 psig for 15 seconds.

C. CORRECT - Electric pump starts if fire protection header is < 140 psig for 2 seconds, D. INCORRECT - Plausible if unsure of pressure setpoints or value of time delay pick up.

Question ID: CM-0628A External Topic ID: AS17: Fire Water Protection Level/Difficulty: F/3 Comments:

REFERENCE:

RO-C-AS17 Note:

FIRE PUMP AUTO START IF FIRE PROTECTION HEADER IS :

1) < 140 PSIG FOR >2 SECONDS...

• ELECTRIC PUMP STARTS from ZPS-413

2) < 130 PSIG FOR >15 SECONDS

• EAST DIESEL PUMP STARTS from ZPS-411

3) < 120 PSIG FOR >25 SECONDS

• WEST DIESEL PUMP STARTS from ZPS-415 DC COOK OPS Page: 130 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 086000 A1.01 Fire Protection system Ability to predict and / or monitor changes in parameters (to prevent exceeding design limits) and associated with operating the Fire Protection System controls including:

Fire Header pressure RO - 2.9 SRO - 3.3 CFR - 41.5 / 45.5 K/A Match Candidate is required to demonstrate ability to predict the fire pump starts based on system header pressure.

Associated objective(s):

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 131 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 66 ID: CM-3153 Points: 1.00 Given the following:

• You are the oncoming Reactor Operator

• This is your first Control Room watch since attending training the previous week Which ONE of the following correctly states the minimum number of days of Control Room Narrative Log entries that must be reviewed prior to taking the watch?

A. 1 B. 3 C. 5 D. 7 Answer: B Answer Explanation:

A. INCORRECT - See B below B. CORRECT- OHI-4000, Conduct of Operations: Standards Attachment 15 states under Log Review: Prior to taking the watch the control room narrative log for the last three days, or since the last time relieving that watch station, whichever is shorter, is the requirement. Since you were attending training the previous week the shortest allowed time to review is three days C. INCORRECT - See B above D. INCORRECT - See B above Question ID: CM-3153 External Topic ID: ADMN-08: Watch Station Turnover Comments:

REFERENCE:

OHI-4000, Conduct of Operations: Standards Attachment 15 DC COOK OPS Page: 132 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 194001 2.1.3 Conduct of Operations Knowledge of shift or short term relief turnover practices RO - 3.7 SRO - 3.9 CFR - 41.10 / 45.13 K/A Match Candidate is required to demonstrate knowledge of the shift turnover requirements.

Associated objective(s):

(RO-C-ADM08-E4) Determine the documents that are required to be reviewed prior to the face to face discussion.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 133 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 67 ID: CM-1319 Points: 1.00 Annunciator 211, Drop 48, PPC-RMS U1 CT ALARM OR ABNORMAL, and Drop 49, PPC-RMS U2 CT ALARM OR ABNORMAL, are received.

• The RO in Unit 2 calls up the RMS Annunciator Status Tree and observes that ERA-7508 is magenta.

• When the RMS CT is checked the ERA-7508 Aux Building EL. 573' Passageway channel is WHITE.

• Radiation Protection has been informed of the status and is taking appropriate actions.

The correct action for the control room operators to take is:

A. remove channel ERA-7508 from scan on the PPC per 1-OHP-4024-211 Attachments 1 &

2.

B. enable the audible alarm on the RMS CT so other channels can be monitored.

C. verify that RP evaluates and enters new Alert and High alarm setpoints.

D. perform event initiated surveillances for PPC failure per PMP-4030-EIS-001.

Answer: A Answer Explanation:

A. CORRECT - The channel indicating magenta on the PPC and White on the RMS CT indicates it has failed. It should be removed from scan so the Control Board PPC alarm can be cleared and functional for monitor other channels.

B. INCORRECT - The PPC is functioning correctly. There is no reason to perform EIS for the ppc.

C. INCORRECT - this failure indicates the channel has failed and does not need the alarm setpoints adjusted.

D. INCORRECT - The Audible would be enabled if the channel was in alert or high alarm so other channels could be monitored. This channel has failed.

Question ID: CM-1319 Comments:

REFERENCE:

02-OHP 4024.211.049 12-OHP-4024-139 Drop 15 DC COOK OPS Page: 134 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 194001 2.1.19 Conduct of operations Ability to use plant computers to evaluate system or component status.

RO - 3.9 SRO - 3.8 CFR - 41.10 / 45.12 K/A Match Candidate is required to demonstrate the ability to use indications including PPC to determine a failed channel and the correctly defeat the channel so the PPC remains able to monitor other points.

Associated objective(s):

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 135 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 68 ID: NRCAUDIT07-0354A Points: 1.00 Unit 2 was operating at 100%.

Given the following events and conditions:

• At 1300 the reactor tripped due to a break outside containment.

• At 1310 the crew enters 2-OHP-4023-ECA-1-1, Loss of Emergency Coolant Recirculation.

• At 1320 the current conditions are:

- RCS pressure is 1050 psig.

- 1 SI pump running, indicating 310 gpm.

- 1 CCP running, indicating 340 gpm.

- Both RHR pumps are off.

- No CTS pump is running.

- Subcooling is 35°F.

- The crew is performing Step 13b.

Which ONE of the following actions, if any, should be taken to establish the required minimum Safety Injection (SI) flow?

REFERENCE PROVIDED A. Minimum flow is 260 gpm, stop the running SI pump and throttle the BIT Discharge Valves.

B. Minimum flow is 320 gpm, stop the running SI pump.

C. Minimum flow is 500 gpm, neither pump may be secured at this time but the BIT Discharge Valves may be throttled.

D. Minimum flow is 600 gpm, neither pump may be secured at this time but the BIT Discharge Valves may be throttled.

Answer: C Answer Explanation:

A. INCORRECT - This is the 200 minute number. (time since trip but applied to graph incorrectly since graph starts at 10).

B. INCORRECT - This is the 100 minute number (time since procedure entry but applied to graph incorrectly since graph starts at 10).

C. CORRECT - Time after trip is 20 minutes, graph starts at 10 minutes, and flow required is 500 gpm. The CCP is providing 340 gpm, and the SI pump 310 - BIT valves will be throttled.

DC COOK OPS Page: 136 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. INCORRECT - This is the 10 minute number (time since procedure entry).

Question ID: NRCAUDIT07-0354A Level/Difficulty: H/3 Comments:

REFERENCE:

2-OHP-4023-ECA-1-1 Reference Provided: 2-OHP-4023-ECA-1-1 Step 13 (page 18-19) and 2-OHP-4023-ECA-1-1 Figure 1 are to be provided.

K/A 194001 2.1.25 Conduct of Operations Ability to interpret reference materials such as graphs, curves, tables, etc.

RO - 3.9 SRO - 4.2 CFR - 41.10 / 43.5 / 45.12 K/A Match Candidate is required to demonstrate the ability to use graphs to determine the correct safety Injection flow.

Associated objective(s):

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 137 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 69 ID: NRCAUDIT07-1010 Points: 1.00 The following conditions exist on Unit 1:

• MODE 3 at normal operating temperature and pressure, preparing for a reactor startup.

• The reactor coolant system has been diluted to the estimated critical boron concentration.

• 1-CRV-470, (Letdown Temperature Control Valve) is in MANUAL.

• All other controls are in AUTOMATIC and functioning NORMALLY.

• Letdown flow is raised from 75 gpm to 120 gpm.

Assuming NO other manipulations, over the next hour, Source Range counts will:

A. RISE due to warmer RCS water exiting the letdown heat exchanger.

B. RISE due to cooler RCS water exiting the letdown heat exchanger.

C. LOWER due to warmer RCS water exiting the letdown heat exchanger.

D. LOWER due to cooler RCS water exiting the letdown heat exchanger.

Answer: C Answer Explanation:

A. INCORRECT - Warmer water exiting the letdown hx and entering the demineralizers causes boron to be released rather than absorbed which will cause source counts to lower. If the RCS was warmer SR counts may indicate slightly higher since the shielding water is less dense.

B. INCORRECT - Warmer water will exit the letdown hx. This would be correct SR response for cooler HX water.

C. CORRECT - With more RCS flow through the letdown Hx and constant CCW cooling flow, the RCS water is not cooled as much. Therefore the inlet temperature of water flowing through the letdown demineralizers is warmer which result in more boron being released into CVCS. This boron once completing the flowpath back into the RCS will result in a drop in source range count rate.

D. INCORRECT - Warmer water will exit the letdown hx. If the RCS was cooler SR counts may indicate slightly lower since the shielding water is more dense.

DC COOK OPS Page: 138 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-1010 Comments:

REFERENCE:

01-OHP-4021-003.001 K/A 194001 2.2.1 Equipment Control Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

K/A Match Candidate is required to demonstrate ability to determine the importance of using procedural guidance to adjust letdown cooling to avoid reactivity effects.

Associated objective(s):

(RO-C-00300-E18) Discuss the following industry experience(s) associated with operation of the CVCS. Include in the discussion: the cause of the event, the impact of the event on plant operations, and any Cook Nuclear Plant specific remedial actions performed to prevent future occurrences of the event:

a. IE Notice 84-70 Reliance on Water Level Instrumentation with a Common Reference Leg

b. SOER 97-01 Potential Loss of High Pressure Injection and Charging Capability from Gas Intrusion

c. SER 08-91 Unexpected Reactivity Additions Due to Boron Dilution While Placing Primary Demineralizers in Service

d. CR 04120060, lifting of letdown safety relief valve, 2-SV-51, while placing 75 gpm letdown orifice in service.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 139 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 70 ID: RO-C-TS01-E8-01 Points: 1.00 Which set of parameters or conditions (from the list below) are used to determine the Operational Modes as defined in Technical Specifications?

1. Auctioneered Low Reactor Coolant Temperature (oF)

2. Reactor Power (Percent Nuclear Instrument)

3. Average Reactor Coolant Temperature (oF)

4. Rated Thermal Power (Percent Thermal)

5. Shutdown Margin (Percent Delta-K/K)

6. Reactivity Condition (K-effective)

A. 1, 2, 5 B. 1, 4, 5 C. 2, 3, 6 D. 3, 4, 6 Answer: D Answer Explanation:

REFERENCE:

Tech Specs Section 1.1, Table 1.1-1 Question ID: RO-C-TS01-E8-01 External Topic ID: TS01-1: TS01-1, Technical Specifications/TRM - LCO/TRO Recognition Level/Difficulty: 3.6 Comments:

REFERENCE:

Tech Specs Section 1.1, Table 1.1-1 K/A 194001 2.2.35 Equipment Control Ability to determine Technical Specification Mode of operation RO - 3.6 SRO - 4.5 CFR - 41.7 / 41.10 / 43.5 / 45.13 K/A Match Candidate is required to demonstrate ability to identify parameters used by Technical Specifications to define operational Mode.

DC COOK OPS Page: 140 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Associated objective(s):

(RO-C-TS01-E8) State three (3) parameters or conditions used in Technical Specifications to determine Operational Modes.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 141 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 71 ID: NRCAUDIT07-0845 Points: 1.00 The following plant conditions exist:

• A Site Area Emergency has been declared due to a Loss of Coolant Accident (LOCA) outside containment.

• A pathway to the environment exists.

• Limited makeup to the Refueling Water Storage Tank is available.

• An EMT must be sent in to provide lifesaving aid to a down operator

• The EMTs current TEDE dose is as follows:

• Exposure for the year: 200 mrem

• Current total lifetime exposure: 1200 mrem Based on these conditions, what is the maximum emergency exposure this EMT may receive while performing this action?

A. 1800 mrem TEDE B. 9800 mrem TEDE C. 23800 mrem TEDE D. 25000 mrem TEDE Answer: D Answer Explanation:

A. INCORRECT - Plausible since this dose would bring the workers dose to 2 Rem which is the maximum ADL TEDE dose allowed per year.

B. INCORRECT - Plausible since this dose would bring the workers dose to 10 Rem which is the maximum for equipment protection in emergency situation.

C. INCORRECT - Plausible since this would bring the worker dose to 25 R for the year, which is the emergency limit, but this limit is independent of previous dose.

D. CORRECT - Emergency exposure limits for life saving or protection of large populations is 25 Rem.

Question ID: NRCAUDIT07-0845 External Topic ID: GFRP: Radiation Protection Level/Difficulty: F/2 Comments:

Reference:

RMT 2080-TSC-001 Section 1.8.2.a DC COOK OPS Page: 142 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 194001 2.3.4 Radiation Control Knowledge of radiation limits under normal or emergency conditions RO - 3.2 SRO - 3.7 CFR - 41.12 / 43.4 / 45.10 K/A Match 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 (RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 143 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 72 ID: RO26-0039 Points: 1.00 Which ONE of the following describes the Operation of the Containment Purge System (in Ventilation Mode) while the Containment equipment Hatch is open?

A. Air flow must be OUT of Containment to minimize radiation levels.

B. Air flow must be INTO Containment to minimize the spread of contamination.

C. Containment Purge Exhaust and Supply flows must be matched to ensure the Containment and Aux Building are maintained at the same pressure.

D. Containment Purge Exhaust flow must equal Supply flow to prevent Ice Condenser doors from opening.

Answer: B Answer Explanation:

A. INCORRECT - Air Flow is maintained into Containment.

B. CORRECT - A lower pressure in containment with respect to the Aux Building will cause airflow into containment and help to minimize the spread of contamination.

C. INCORRECT - Aux Building and Containment Pressures are not maintained equal.

D. INCORRECT - This concern is addressed by the Upper/Lower Containment Pressure balance. Air flow must be out of containment.

Question ID: RO26-0039 Comments:

REFERENCE:

01-OHP-4021-028-005, Operation of The Containment Purge System, Attachment 2, step 3.7 and Figure 2, K/A 194001 2.3.13 Radiation Control Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

RO - 3.4 SRO - 3.8 CFR - 41.12 / 43.4 / 45.9 / 45.10 DC COOK OPS Page: 144 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A Match Candidate is required to demonstrate knowledge of radiological safety precautions within the Containment Purge procedure regarding operation of the fans during refueling outage.

Associated objective(s):

(RO-C-02800-E4) State the basis for the difference in supply and exhaust flowrates between the containment upper volume and containment lower volume during purge operations.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 145 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 73 ID: NRC2010-74 Points: 1.00 Given the following conditions on Unit 1:

• A reactor trip occurred from 15% power.

• Safety Injection was NOT actuated and was NOT required.

• 1-OHP-4023-E-0, Reactor Trip or Safety Injection, has been performed, and a transition to 1-OHP-4023-ES-0.1, Reactor Trip Response, has been made.

The following conditions exist:

• Tavg is STABLE at 547oF

• Pressurizer level is 11% and lowering slowly

• RCS subcooling is 32oF and lowering slowly

• All NR SG levels are 28 - 30%,

• AFW flows indicate 0 pph

• Containment pressure is 0.7 psig and rising slowly Which ONE of the following describes the appropriate actions for these conditions?

A. Actuate SI and return to 1-OHP-4023-E-0 step 1.

B. Go to 1-OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink.

C. Go to 1-OHP-4023-FR-I.2, Response to Low Pressurizer Level.

D. Manually start ECCS pumps and continue with 1-OHP-4023-ES-0.1.

Answer: A Answer Explanation:

A. CORRECT - The foldout page directs an SI and return to 1-OHP-4023-E-0 if subcooling is

< 40oF.

B. INCORRECT - SG levels are adequate for Heat Sink even with no AFW flow. Plausible due to loss of subcooling the student could want to increase cooling to the S/G as well as missing the Narrow Range level.

C. INCORRECT - Initiating SI and returning to 1-OHP-4023-E-0 would take precedence over 1-OHP-4023-FR-I.2. Plausible due to the desire of the student to recover PZR level.

D. INCORRECT - This may be the action required in other emergency procedures (ES-1.2) but a return to 1-OHP-4023-E-0 is required to verify proper alignment of ECCS equipment. ES-0.1 is considered a 'non-accident' EOP. Plausible due to this action being elsewhere in the EOP network.

DC COOK OPS Page: 146 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRC2010-74 Level/Difficulty: F/3 Comments:

REFERENCE:

1-OHP-4021-ES-0.1 (Foldout Page)

K/A 194001 2.4.2 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 K/A Match Candidate is required to demonstrate knowledge of conditions (setpoints) that would require a transition from Reactor Trip Response to re-entry into the Reactor Trip/SI EOP.

Associated objective(s):

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 147 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 74 ID: CM-2925 Points: 1.00 Step 9 of OHP-4023-E-1, Loss of Reactor or Secondary Coolant, checks RCS and SG pressures. If the pressure in any SG is lowering or RCS pressure is rising, the operators are returned to step 1.

Which ONE of the following statements is true regarding this step?

A. The concern addressed by this step is the presence of a secondary side break in which the faulted SG is still depressurizing in an uncontrolled manner. If this is the case the operator should proceed to ES-1.2, Post LOCA Cooldown And Depressurization.

B. If the operator has completed the required actions of E-2 to isolate the faulted SG(s), and the pressure in the faulted SG(s) is still lowering, the operator will have to return to E-2 to continue attempts to isolate the faulted SG(s).

C. This procedure is used to recover from both a LOCA and secondary side break. Both of these conditions being met indicates that both events are occurring and recovery must continue with more restrictive SI termination criteria.

D. Eventually, if a faulted SG exists, it will blow down to atmospheric pressure and dry out, then RCS pressure will stabilize or rise and the SI termination criteria of this procedure should be met.

Answer: D Answer Explanation:

A. INCORRECT - Plausible because a faulted SG which is continuing its blowdown is one of the concerns addressed by the step, but INCORRECT because the transition will be to ES-1.1 SI Termination when the depressurization is complete.

B. INCORRECT - Plausible because a faulted SG is addressed by E-2; INCORRECT, because, once E-2 is completed, it does not require re-entry (for the same SG).

C. INCORRECT - Plausible if the symptoms of a faulted SG used in Step 9 are assumed to always be due to a concurrent accident.

D. CORRECT - Completion of blowdown in the faulted SG will result in meeting conditions for SI termination, a continuous action step in E-1.

Question ID: CM-2925 External Topic ID: PROC-4: PROC-4, Procedure Step Basis Comments:

REFERENCE:

OHP-4023-E-1, Loss of Reactor or Secondary Coolant DC COOK OPS Page: 148 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 194001 2.4.6 Emergency Procedures / Plan Knowledge of EOP mitigation strategies RO - 3.7 SRO - 4.7 CFR 41.10 / 43.5 / 45.13 K/A Match Candidate is required to demonstrate knowledge of a critical decision point within E-1 to determine (Mitigation Strategy) correct path to cooldown and depressurize the plant.

Associated objective(s):

(RO-C-EOP09-E41) For each of the E-1 Series procedures, describe any Procedural Loops.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 149 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 75 ID: RO-C-ADM02-E25-3 Points: 1.00 Which ONE of the following describes the actions required if the 23 Annunciator Visual Display Units (VDUs) quit functioning?

A. Immediately trip the reactor and enter OHP-4023-E-0, reactor Trip Response.

B. Assign an operator to monitor the BOP Annunciator Workstation, begin a Rapid Power Reduction, and contact the Annunciator System Administrator to troubleshoot and repair the issue.

C. Stabilize the plant, cycle the VDU and contact the Annunciator System Administrator to troubleshoot and repair the issue.

D. Trip the reactor and enter OHP-4023-E-0, reactor Trip Response ONLY if the BOP AND US Annunciator Workstations are ALSO Failed.

Answer: C Answer Explanation:

A. INCORRECT - It is desirable to maintain the plant stabile while no annunciators are available.

B. INCORRECT - The BOP and US desk may provide backup monitoring but the plant should be maintained stable even if the whole system is down.

C. CORRECT - It is desirable to maintain the plant stabile while no annunciators are available. Monitoring the control boards with a stable plant is the safest.

D. INCORRECT - The BOP and US desk may provide backup monitoring but the plant should be maintained stable even if the whole system is down. The E-Plan Classification is raised if there is a transient in progress.

Question ID: RO-C-ADM02-E25-3 Level/Difficulty: F/3 Comments:

REFERENCE:

2-OHP-4024-212 Drop 39 K/A 194001 2.4.32 Emergency Procedures/Plan Knowledge of operator response to loss of all annunciators.

RO - 3.6 SRO - 4.0 CFR - 41.10 / 43.5 / 45.13 DC COOK OPS Page: 150 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A Match Candidate is required to demonstrate knowledge of the response required for a loss of all annunciators.

Associated objective(s):

(RO-C-ADM02-E25) Describe how to properly respond to Control Room alarms in accordance with OHI-4000, "Conduct of Operations: Standards," including proper communications and required log keeping.

(RO31 NRC Exam 2018 51-75)

DC COOK OPS Page: 151 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 76 ID: CM-40207 Points: 1.00 SRO Only Unit 2 was operating at 40% power and experienced a severe Feedwater Break. SG 22 has completely depressurized and 2-OHP-4023-E-2, Faulted Steam Generator Isolation, has been entered.

The following conditions exist:

• RCS Tcolds are 500oF and slowly lowering.

• All Main Feedwater Isolation valves are closed.

• All SG Stop valves and Stop Valve Dump valves are closed.

• Pressure in SGs 21, 23, and 24 are lowering in an uncontrolled manner.

• STEAM GEN 1, 3, and 4 SF > FWF FLOW MISMATCH annunciators are LIT.

• SG 1, 2, 3, 4 STEAM LINE PRESSURE LOW annunciator just alarmed.

Which ONE of the following procedural transitions, if any, is required based on these conditions?

A. 2-OHP-4023-ECA-2.1, Uncontrolled Depressurization of all Steam Generators.

B. 2-OHP-4023-FR-H.1, Response to Loss of Secondary Heat Sink.

C. Do NOT transition, remain in 2-OHP-4023-E-2, Faulted Steam Generator Isolation.

D. 2-OHP-4023-FR-H.5, Response to Steam Generator Low Level.

Answer: A Answer Explanation:

A. CORRECT - 2-OHP-4023-E-2, Step 3 requires a transition to ECA-2.1 in the event that no SG Pressure Boundary is intact. SG 21, 23, and 24 Steam Gen Steam Line Pressure Low annunciators are set at 500 psig which equates to 470 F Tsat.

These indications show that the pressure reduction in the SGs is from the cooldown of the SGs rather than the RCS cooldown.

B - INCORRECT - No failure of AFW is indicated, adequate AFW flow should be assumed to exist. No reason to enter FR-H.1.

C - INCORRECT - A Transition is required since all SG pressures are lowering.

D - INCORRECT - FR-H.5 is a yellow path procedure. Yellow path procedures are never required to be entered.

DC COOK OPS Page: 152 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: CM-40207 External Topic ID: PROC-2: Procedure Entry Conditions Comments:

REFERENCE:

2-OHP-4023-E-2, Faulted Steam Generator Isolation K/A 000007 2.4.45 Reactor Trip - Stabilization Ability to prioritize and interpret the significance of each annunciator or alarm RO - 4.1 SRO - 4.3 CFR - 41.10 / 43.5 / 45.3 / 45.12 K/A Match Candidate is required to demonstrate knowledge of the significance of the steam pressure alarms as they relate to the RCS and the required procedure actions.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO-C-EOP07-E17) For E-2 and ECA-2.1 identify the Procedure Transitions.

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 153 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 77 ID: NRCAUDIT07-0993A-SRO Points: 1.00 SRO Only The first charging pump is shutdown during safety injection termination per OHP-4023-ES-1.1, Safety Injection Termination, following a Unit 2 LOCA.

The following parameter trends are noted:

• Subcooling is +41°F and lowering.

• Pressurizer level is 25% and lowering.

• RCS pressure is 1680 psig and lowering.

Which ONE of the following describes parameter(s) that were affected by the charging pump shutdown and what procedure will be used to stabilize the plant?

Lower injection flow is causing:

A. pressure to lower. Go to OHP-4023-ES-1.2, Post LOCA Cooldown and Depressurization.

B. pressure to lower. Go to OHP-4023-FR-C.2 Response to Degraded Core Cooling.

C. temperature to rise. Go to 2-OHP-4023-ES-1.2, Post LOCA Cooldown and Depressurization.

D. temperature to rise. Go to OHP-4023-FR-C.2 Response to Degraded Core Cooling.

Answer: A Answer Explanation:

A. CORRECT - After the first CCP is stopped, RCS pressure is checked to determine if 1 CCP will be sufficient to maintain control with the lower flow. If the leak is large enough to cause RCS pressure to lower, then a transition is made to ES-1.2 to address the leak.

B. INCORRECT - A transition is made to ES-1.2 since RCS Pressure is lowering. FR-C.2 Entry requires a Lower RVLIS Level (PZR is 25%). Plausible since subcooling is lowering.

C. INCORRECT - The primary concern is lowering pressure and level. The transition is made based on pressure.

D. INCORRECT - A transition is made to ES-1.2 since RCS Pressure is lowering. FR-C.2 Entry requires a Lower RVLIS Level (PZR is 25%). Plausible since subcooling is lowering.

DC COOK OPS Page: 154 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0993A-SRO External Topic ID: PROC-2: Procedure Entry Conditions Level/Difficulty: H/3 SRO ONLY Comments:

REFERENCE:

12-OHP-4023-ES-1.1 Background Step 5 K/A 000009 EA2.15 Small Break LOCA Ability to determine and interpret the following as they apply to a small break LOCA:

RCS parameters RO - 3.3 SRO - 3.4 CFR - 41.7 / 41.10 / 43.5 / 45.13 KA Match Candidate is required to demonstrate the ability to determine required procedural actions based on RCS parameters during a small break LOCA SRO - 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. [55.43(b)(5)]

Associated objective(s):

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

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 155 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 78 ID: RO-C-EOP03-E14-1-SRO Points: 1.00 SRO Only Given the following conditions on Unit 1:

• A power escalation is in progress.

• At approximately 35% power the Main Turbine Trips.

The following conditions now exist:

• All 4 Turbine Stop Valve Closed Status Lights are LIT.

• All Rod Position Indicators (RPIs) have lost power.

• Train A Reactor Trip Breaker is OPEN.

• Train B Reactor Trip Breaker is CLOSED.

• Wide Range (WR) Log Power = 7%.

• WR Startup Rate = 0.0 DPM and stable.

Which ONE of the following actions is required?

A. Implement 1-OHP-4023-E-0, Reactor Trip or Safety Injection. Following completion of Immediate Actions, transition to 1-OHP-4023-ES-0.1, Reactor Trip Response.

B. Implement 1-OHP-4023-E-0, Reactor Trip or Safety Injection. During verification of Reactor Trip, transition to 1-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS.

C. Implement 1-OHP-4022-001-002, Loss of Load (Load Rejection). When directed go to 1-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS.

D. Implement 1-OHP-4022-001-002, Loss of Load (Load Rejection). Upon Turbine Trip verification, go to 1-OHP-4023-E.0, Reactor Trip or Safety Injection.

Answer: B Answer Explanation:

A. INCORRECT - A transition to FR-S.1 is required. Plausible as E-0 is the correct entry procedure and student may consider actions in E-0 capable of resolving the problem.

B. CORRECT - Due to flux being greater than 5% and not lowering, E-0, Step 1 RNO will require a transition to FR-S.1. Since a Reactor trip cannot be verified (flux NOT lowering), manual control rod insertion will be required.

C. INCORRECT - E-0 should be implemented since a Turbine Trip greater than P-8 should result in a Reactor trip. Plausible as student may key in on current power level and go to procedure governing for a loss of load with a reactor power level below the reactor trip setpoint.

DC COOK OPS Page: 156 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. INCORRECT - E-0 should be implemented since a Turbine Trip greater than P-8 should result in a Reactor trip. E-0 takes precedence over AOPs. Plausible as student may key in on current power level and go to procedure governing for a loss of load with a reactor power level below the reactor trip setpoint.

Question ID: RO-C-EOP03-E14-1-SRO External Topic ID: PROC-2: Procedure Entry Conditions Level/Difficulty: H/2 SRO ONLY Comments:

REFERENCE:

1-OHP-4023-E-0 Step 1 K/A 000029 2.4.4 Anticipated Transient Without Scram (ATWS)

Emergency Procedures/Plan Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures RO - 4.5 SRO - 4.7 CFR - 41.10 / 43.2 / 45.6 K/A Match Candidate is required to demonstrate the ability to recognize plant conditions and select the appropriate procedure flow path.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO-C-EOP03-E14) List the Immediate Operator Actions for E-0.

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 157 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 79 ID: RO-C-EOP-E10-1A Points: 1.00 SRO Only Unit 1 reactor has been manually tripped due to a steam generator tube rupture.

1-OHP-4023-E-3 has been performed and a transition made to 1-OHP-4023-ES-3.1, Post-SGTR Cooldown using Backfill. The STA has identified a YELLOW path on the Heat Sink Status Tree for steam generator pressure.

The crew has entered 1-OHP-4023-FR-H.2, Response to Steam Generator Overpressure.

Given the following plant conditions:

• Steam Generator #13 Pressure - 1120 psig

• Steam Generator #13 NR Level - 92%

Which ONE of the following is the action relative to steam release in accordance with FR-H.2?

A. Steam may be released through the TDAFP Steam Supply Valve ONLY since the steam is contaminated.

B. Steam release should NOT occur because it may result in two phase flow and water hammer, potentially damaging pipes and valves.

C. Steam release should NOT occur because the higher pressure is desired to aid in the Backfill cooldown.

D. Steam may be released without restriction since narrow range level has been adequately established and it provides the fastest method to cooldown the SG.

Answer: B Answer Explanation:

A. INCORRECT - The TDAFP would be the preferred option if the level was <91 %.

B. CORRECT - Per 1-OHP-4023-FR-H.2 Step 3, if the SG level is >91%, a transition is made to 1-OHP-4023-FR-H.3 to address the high level that may be causing the pressure concern. Also 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 - With Steam Pressure this high, it is desired to lower pressure to prevent piping failure.

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

DC COOK OPS Page: 158 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-EOP-E10-1A 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 K/A 000038 EA2.16 Steam Generator Tube rupture (SGTR)

Ability to determine and interpret the following as they apply to a SGTR:

Actions to be taken if S/G goes solid and water enters steam line RO - 4.2 SRO - 4.6 CFR - 41.7 / 41.10 / 43.5 / 45.13 K/A Match Candidate is required to demonstrate the ability to determine correct actions to take if water enters the steam line SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 159 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 80 ID: RO-C-0820360101-E3-1 Points: 1.00 SRO Only Given the following plant conditions:

• Unit 2 was operating at 100% power when a loss of CRID 1 occurs Which ONE of the following describes the correct Operator response?

A. Perform 2-OHP-4021-082-008, Operation of CRID Power Supplies to restore ESFAS functions as quickly as possible. Prepare to enter 2-OHP-4022-001-006 Rapid Power Reduction Response if the CRID can NOT be restored.

B. Enter 2-OHP-4022-001-006 Rapid Power Reduction Response due to loss of ESFAS functions. Steps from 2-OHP-4021-082-008, Operation of CRID Power Supplies, may be performed concurrently with 2-OHP-4022-001-006.

C. Enter 2-OHP-4023-E-0, Reactor Trip or Safety Injection.

Steps from 2-OHP-4021-082-008, Operation of CRID Power Supplies, may be performed concurrently with 2-OHP-4023-ES-0.1, Reactor Trip Response.

D. Enter 2-OHP-4023-E-0, Reactor Trip or Safety Injection.

After exiting 2-OHP-4023-E-0 perform 2-OHP-4021-082-008, Operation of CRID Power Supplies to restore ESFAS functions as quickly as possible.

Perform 2-OHP-4023-ES-0.1, Reactor Trip Response after the CRID has been restored.

Answer: C Answer Explanation:

A. INCORRECT - The Reactor will trip due to loss of CRID 1. ESFAS function is lost for 1 train and 2-OHP-4021-082-008 is used to restore power to CRID and determine equipment lost.

B. INCORRECT - The Reactor will trip due to loss of CRID 1. ESFAS function are lost for 1 train but immediate shutdown is not required.2-OHP-4021-082-008 is used to restore power to CRID and determine equipment lost.

C. CORRECT - The reactor will trip and immediate actions should be performed.

2-OHP-4021-082-008 can be performed concurrently with 2-OHP-4023-ES-0.1. OHI-4023, Abnormal/Emergency Procedure User's Guide allows Procedures to be implemented concurrently with Emergency Procedures after the immediate actions are complete.

D. INCORRECT - ES-0.1 is still part of the emergency procedure network and so should be performed with 2-OHP-4021-082-008.

DC COOK OPS Page: 160 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-0820360101-E3-1 Level/Difficulty: H/3 Comments:

REFERENCE:

02-OHP-4021-082-008 Operation of CRID Power Supplies, Table 1 OHI-4023, Abnormal/Emergency Procedure User's Guide K/A 000057 2.4.8 Loss of Vital AC Electrical Instrument Bus Knowledge of how abnormal operating procedures are used in conjunction with EOPs RO - 3.8 SRO - 4.5 CFR: 41.10 / 43.5 / 45.13 K/A Match Candidate is required to demonstrate the knowledge on the analysis of plant status and coordination of the required procedures.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO-C-0820360101-E3) explain the procedural mitigation strategy for Loss of CRID Instrument Inverters and Power Panels (RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 161 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 81 ID: RO-C-EOP09-E35-2 Points: 1.00 SRO Only Given the following conditions:

• Unit 1 experienced a Large Break Loss of Coolant Accident about 30 minutes ago.

• Emergency Core Cooling System (ECCS) Pumps are running in Cold Leg Recirculation.

• Transfer of ECCS Pumps to recirculation is complete.

The East and WEST RHR Pump amps and flows have started to oscillate.

Which of the following lists the required transition and subsequent actions that will be taken?

A. Transition to 1-OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation Procedure and shutdown all pumps taking suction from the recirculation sump.

B. Transition to 1-OHP-4023-ECA-1.3, Sump Blockage Control Room Procedure and shutdown all pumps taking suction from the recirculation sump.

C. Transition to 1-OHP-4023-ECA-1.1, Loss of Emergency Coolant Recirculation Procedure and shutdown 1 CCP, SI, and RHR pump and both CTS pumps.

D. Transition to 1-OHP-4023-ECA-1.3, Sump Blockage Control Room Procedure and shutdown 1 CCP, SI, and RHR pump and both CTS pumps.

Answer: D Answer Explanation:

A. INCORRECT - ECA-1.1 is generally used if recirculation is compromised, but this indicates a blockage. All pumps taking suction from a LOW RWST are stopped, so this is a plausible action.

B. INCORRECT - correct procedure transition. All pumps taking suction from a LOW RWST are stopped, so this is a plausible action.

C. INCORRECT - ECA-1.1 is generally used if recirculation is compromised, but this indicates a blockage. The flow reduction is correct.

D. CORRECT - Per the ES-1.3, Foldout Page and step 6.u, transition is made if both trains are indicating cavitation. Steps 1-5 of ECA-1.3 stops 1 train of ECCS and both CTS pumps.

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-EOP09-E35-2 Level/Difficulty: H/3 Comments:

REFERENCE:

OHP-4023-ES-1.3 step 6, Foldout Page item #2, OHP-4023-ECA-1.3 Steps 1-5 K/A 00WE11 EA2.1 Loss of Emergency Coolant Recirculation Ability to determine and interpret the following as they apply to the Loss of Emergency Coolant Recirculation:

Facility conditions and selection of appropriate procedures during abnormal and emergency operations RO - 3.4 SRO - 4.2 CFR - 41.7 / 41.10 / 43.5 / 45.13 K/A Match Candidate is required to demonstrate knowledge of the conditions and selection of procedures related to Loss of Recirculation.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO-C-EOP09-E35) For each of the E-1 Series procedures, discuss the basis or reason for all Foldout Page Items (RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 163 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 82 ID: NRCAUDIT07-0860-SRO Points: 1.00 SRO Only The following conditions exist on Unit 1:

• Unit is heating up in Mode 3 following a Refueling Outage.

• An AEO reports that the boric acid filter DP is pegged HIGH.

• Discharge pressure on the #1 Boric Acid Transfer (BAT) Pump is pegged HIGH.

Which ONE of the following describes the effect on the operability of the boron injection flowpath from the Boric Acid Storage Tank (BAST) and the actions required?

The BAST to Chemical Volume Control System (CVCS) Suction Boration Flowpath is:

A. INOPERABLE. Direct actions to bypass the filter.

B. INOPERABLE. Direct actions to align the standby filter.

C. OPERABLE. Brief the RO to use the #2 BAT pump which is lined up to the Middle BAST if Emergency Boration is required.

D. OPERABLE. Brief the RO to use 1-CS-294, Alternate Boration to CVCS Charging Pumps Suction Header Inlet Valve if Emergency Boration is required.

Answer: A Answer Explanation:

A. CORRECT - The Boric Acid filter may be bypassed so it may deliver the required flow and therefore is Inoperable until aligned.

B. INCORRECT - No Standby Filter. Plausible because it is Inoperable and several systems have standby filters.

C. INCORRECT - Plausible since the other pump/tank combination may have a separate flowpath except #2 pp must also use the clogged filter.

D. INCORRECT - Plausible since this is a boric acid flowpath however, this flowpath does not deliver required flow + downstream of filter.

DC COOK OPS Page: 164 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: Question ID:

External Topic ID: 0300-2: CVCS Makeup System Level/Difficulty: H/3 Comments:

REFERENCE:

1-OHP-4030-105-002V Lineup sheet 9, TRM 8.1.1 K/A 000024 AA2.02 Emergency Boration Ability to determine and interpret the following as they apply to the Emergency Boration:

When use of manual boration valve is needed.

RO - 3.9 SRO - 4.4 CFR - 41.7 / 41.10 / 43.5 / 45.13 K/A Match Candidate is required to demonstrate ability to determine system operability requirements with regard to the use of various manual flow paths.

SRO - Facility operating limitations in the TS and their bases. [55.43(b)(2)] Also Question has candidate assess the plant conditions and determine the correct actions based detailed procedural steps (RNO). [55.43(b)(5)]

Associated objective(s):

(RO-C-00300-E17) Given a description of plant conditions and/or the results of a Surveillance Test, determine the following: All applicable TS, System Operability, and most limiting LCO.

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 165 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 83 ID: RO-C-01300-E21-2 Points: 1.00 SRO Only The following events occur during a power escalation on Unit 2:

• Intermediate Range Channel N-35 failed when power was 6%. It is currently Inoperable with all required actions taken per 02-OHP-4022-013-003 INTERMEDIATE RANGE MALFUNCTION

• Power Range (PR) nuclear instrumentation channels all indicate 8%

• Intermediate Range Channel N-36 indication drops to 1e-11 amps Which ONE of the following describes the actions to be directed by the SRO and the correct basis?

A. Raise Power to > 10% since overlap with power range instruments has been verified. The Intermediate Range is not required above 10% since the Intermediate Range Low Power trip is redundant to the Power Range Low Flux trip protection for a Control Rod Withdrawal accident.

B. Verify the Reactor tripped on the loss of 2 intermediate range channels and enter 02-OHP-4023-E-0 Reactor Trip Response. Immediately energize both Source Range Channels because intermediate range channel protection is lost for the Start Up Reactivity Excursion.

C. Stabilize the plant and do NOT make any reactivity changes. Enter Technical Specification 3.0.3 since the plant is outside of the design basis.

D. Immediately suspend positive reactivity changes and reduce power to <P6 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> because no intermediate range instrumentation channels are operable in the event of a Control Rod Withdrawal accident.

Answer: D Answer Explanation:

A. INCORRECT - Positive reactivity changes must be stopped per TS 3.3.1 Action F.1. Power should have been raised to > 10% with 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the first channel failure.

TS. 3.3.1 Action E.2. The Intermediate range does provide redundant protection to the Low Flux Power Range trip so it is not required above 10%.

B. INCORRECT - The reactor would have tripped if IR N36 had failed high. 1e-11 is a low channel failed and N-35 trip is bypassed. SR and IR above P-6 provide Start-Up reactivity protection.

C. INCORRECT - Negative reactivity can be added. This condition is covered in Action F so 3.0.3 does not need to be entered.

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. CORRECT - Positive reactivity changes must be stopped per TS 3.3.1 Action F.1. and a shutdown commenced with 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> since control rod withdrawal protection is lost.

Question ID: RO-C-01300-E21-2 Level/Difficulty: H/3 Comments:

REFERENCE:

Tech Spec. 3.3.1 Actions E & F and Basis for IR trip K/A 000033 2.2.25 Loss of Intermediate Range Nuclear Instrumentation Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits K/A Match Candidate is required to demonstrate knowledge of Tech Spec actions and the bases for them in a Loss of Intermediate Range Nuclear Instrumentation condition.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO-C-01300-E21) Given the plant and NIS in any operating condition:

Determine if LCOs are met Apply action requirements as appropriate (RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 167 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 84 ID: 2007-0463A Points: 1.00 SRO Only Given the following events and conditions:

• A loss of condenser vacuum occurred on Unit 2

• The operators are performing steps of 2-OHP-4022-001-006 Rapid Power Reduction Response.

• Power Range NIs read 27% power.

• Turbine load is 25%.

Which ONE of the following statements describes Turbine vacuum limit and the SRO required directions?

Turbine Operation is NOT allowed at less than:

A. 24.9 inches Hg, trip the turbine then reduce reactor power to 17% and trip the reactor.

B. 24.9 inches Hg, reduce reactor power to 17%, trip the reactor and then trip the turbine.

C. 23.4 inches Hg, trip the reactor and then trip the turbine immediately after reaching 23.4 inches.

D. 23.4 inches Hg, trip the turbine then reduce reactor power to 17% and trip the reactor.

Answer: D Answer Explanation:

A. INCORRECT - 30 minutes of operation is allowed between 24.9 and 23.4.

B. INCORRECT - 30 minutes of operation is allowed between 24.9 and 23.4.

OHP-4022-001-006 has the turbine tripped when below P-8 and then power lowered to 17 % for the reactor trip.

C. INCORRECT - OHP-4022-001-006 has the turbine tripped when below P-8 and then power lowered to 17 % for the reactor trip. The turbine can be tripped and then the reactor shutdown (tripped at <17%)

D. CORRECT - Turbine must not be operated at low powers when <23.4 inches.

OHP-4022-001-006 has the turbine tripped when below P-8 and then power lowered to 17 % for the reactor trip.

DC COOK OPS Page: 168 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: 2007-0463A Comments:

REFERENCE:

02-OHP-4024-MT-DCS Drops 134 - 136, OHP-4022-001-006, TDB-2-FIG-11-19 K/A 000051 AA2.02 Loss of Condenser Vacuum Ability to determine and interpret the following as they apply to the 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 / 45.13 K/A Match Candidate is required to demonstrate the ability to interpret correct actions for when a turbine and reactor trip is required during a Loss of Condenser Vacuum condition.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 169 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 85 ID: RO-C-EC02-E4-2 Points: 1.00 SRO Only Given the following plant conditions:

• Unit 2 is at 100% power when a fire occurs in the Control Room Cable Vault.

• Due to the large amount of smoke in the Control Room, it is decided by the Shift Manager that the main control room must be evacuated.

• Prior to evacuation, the crew enters 02-OHP-4025-001-001, Emergency Remote Shutdown and attempts to trip the reactor in the Control Room.

• The Reactor trip breakers remain closed.

Which ONE of the following responses below describes the proper operating crew actions for the given conditions and describes the method used to monitor the reactor shutdown?

A. Exit the 02-OHP-4025-001-001, Emergency Remote Shutdown procedure and perform reactor trip actions in accordance with 02-OHP-4023-E-0, Reactor Trip or Safety Injection.

Verify Rod Bottom lights lit and WR flux lowering prior to exiting the control room.

B. Exit the 02-OHP-4025-001-001, Emergency Remote Shutdown procedure and perform reactor trip actions in accordance with 02-OHP-4023-FR-S.1, Response to Nuclear Power Generation/ATWS. Verify WR flux lowering prior to exiting the control room.

C. Direct the Reactor Trip, Bypass breakers and MG Output breakers to be opened locally as required. Continue 02-OHP-4025-001-001, Emergency Remote Shutdown actions and evacuate the control room. Direct the STA to verify all rods inserted and WR flux lowering using the PPC connection from Unit 1 Control Room.

D. Direct the Reactor Trip, Bypass breakers and MG Output breakers to be opened locally as required. Continue 02-OHP-4025-001-001, Emergency Remote Shutdown actions and evacuate the control room. Verify WR flux lowering on the LSI panel.

Answer: D Answer Explanation:

A. INCORRECT - Carrying out 02-OHP-4023-E-0 post trip actions is not the correct procedural action path to expeditiously evacuate the MCR.

B. INCORRECT - Carrying out 02-OHP-4023-FR-S.1 trip actions is not the correct procedural action path to expeditiously evacuate the MCR.

C. INCORRECT - The Reactor would be tripped locally but the STA is not directed to monitor the Unit 2 from the PPC. PPC Indications could be wrong due to the fire.

D. CORRECT - RX is tripped from the Control Room or Locally. NRI-23 WR flux is monitored on the LSI Panel.

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Note:

Per OHI-4023:

Section 4.2.3 Implementation of 1(2)-OHP-4025-001-001, Emergency Remote Shutdown, takes precedence over actions specified in other plant procedures. , Section 4.1 When the fire affected units Reactor is tripped per 1(2)-OHP-4025-001-001, Emergency Remote Shutdown, remain in the ERS procedure. Do NOT implement 1(2)-OHP-4023-E-0, Reactor Trip or Safety Injection. Actions performed by the ERS procedure will ensure the ERS Team has control of reactivity, Reactor Coolant System (RCS) temperature and inventory, and establishes a heat sink to stabilize the unit in HOT STANDBY.

Question ID: RO-C-EC02-E4-2 External Topic ID:

Level/Difficulty: H/3 Comments:

REFERENCE:

OHI-4023 Section 4.2.3 & Attachment 3 Section 4.1 K/A 000068 2.4.21 Control room evacuation 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.

K/A Match Candidate is required to demonstrate the correct procedural flow path and implementation and identify the instrumentation that would be used to assess reactivity safety function form outside the control room.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

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

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 171 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 86 ID: CM-7786-SRO Points: 1.00 SRO Only Given the following conditions on Unit 2:

• The unit is in Mode 5 at 150°F with the Reactor Coolant System (RCS) intact.

• Both trains of Residual Heat Removal (RHR) are available, with the West RHR train in operation.

• Both Pressurizer (PRZ) Power Operated Relief Valves (PORVs) are lined up for Low Temperature Operating Pressure (LTOP) mode of operation.

• The West Centrifugal Charging Pump (CCP) and Both Safety Injection (SI) Pumps are Tagged Out.

• Due to indication of cavitation on West RHR Pump, the crew enters 2-OHP-4022-017-001, Loss of RHR Cooling.

Per the procedure, the crew isolates both trains of RHR by placing both RHR pumps in Pull-To-Lock and closes RHR Suction Valves IMO-128 and ICM-129.

How does this action affect the Technical Specification operability of the LTOP System?

A. Technical Specification requirements are met. Both PRZ PORVs are available for overpressure control.

B. Technical Specification requirements are NOT met. The RHR Suction Relief must be available for overpressure protection.

C. Technical Specification requirements are NOT met. RCS temperature is too low.

D. Technical Specification requirements are met. Isolating RHR Suction Valves does NOT affect operability of the RHR Suction Relief.

Answer: A Answer Explanation:

A. CORRECT - Tech. Spec requirements are met. Both PORVs are available for overpressure control. T.S. 3.4.12 requires 2 overpressure protection flow paths whether both PORVs a PORV and RHR suction relief. Closing the RHR suction valves isolates the RHR suction reliefs, but the PORVs are still available.

B. INCORRECT - The RHR suction Relief is not required since 2 PORVs are available and only 1 CCP is available. (True if both CCPs available)

C. INCORRECT - RCS temperature must be > 140°F (old # was 152°F).

D. INCORRECT - The Suction Relief is downstream of the RCS to RHR isolation valves.

DC COOK OPS Page: 172 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: CM-7786-SRO External Topic ID: TS01-1: Technical Specifications/TRM - LCO/TRO Recognition Level/Difficulty: H/3 Comments:

REFERENCE:

Tech Spec 3.4.12 K/A 005000 A2.02 Residual Heat Removal System (RHRS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Pressure transient protection during cold shutdown.

RO - 3.5 SRO - 3.7 CFR - 41.5 / 43.5 / 45.3 / 45.13 K/A Match When presented with a RHR malfunction the candidate is required to demonstrate the ability to predict the impact of over pressure protection capability based on plant alignment and identify the Technical Specification requirements based on the alignment.

SRO - 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). [55.43(b)(2)]

Associated objective(s):

(RO-C-01700-E14) Given a description of plant conditions and TS, determine the following:

All applicable TS, System Operability, and most limiting LCO.

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 173 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 87 ID: 2008NRC-0211 Points: 1.00 SRO Only Given the following plant conditions on Unit 2:

• The Unit is cooling down on RHR.

• RCS Temperature is 340°F and lowering.

• RCS pressure is 300 psig and lowering.

• Pressurizer level is 22% and lowering.

• Containment pressure is 0.1 psig and stable.

• VRS-2505, U2 Vent Noble Gas is in Alarm.

• VRS-2503, U2 Vent Iodine Monitor radiation levels are trending higher.

• SG levels stable at ~ 42%.

Which ONE of the following identifies the problem and the associated action?

A. A LOCA has occurred on the suction of the RHR pump. Enter 2-OHP-4022-002-015, Mode 4 LOCA.

B. LTOP (Low Temperature Over Pressure) actuated and one PORV is stuck open. Enter 2-OHP-4022-002-015, Mode 4 LOCA.

C. Letdown line pressure relief valve has failed open. Enter 2-OHP-4022-002-020, Excessive Reactor Coolant Leakage D. A LOCA has occurred in the area of the Regenerative Heat Exchanger. Enter 2-OHP-4022-002-020, Excessive Reactor Coolant Leakage Answer: A Answer Explanation:

A. CORRECT - During any LOCA, RCS pressure and inventory will fall. Rising indication on the Aux Bldg. Radiation monitors is indicative of the LOCA outside containment (RHR pump suction).

B. INCORRECT - When LTOPs operation results in a PORV opening, RCS pressure will drop, but PZR level should rise due to voiding in the reactor vessel head. Initial conditions do NOT support LTOP auto operation.

C. INCORRECT - Failure of IRV-300 open results in increased diversion of RHR flow (RCS inventory) to letdown but the flow is contained and would not lead to rising Aux Bldg. RMS indications.

D. INCORRECT - This leak is inside containment.

DC COOK OPS Page: 174 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: 2008NRC-0211 Comments:

REFERENCE:

02-OHP-4022-002-015, Mode 4 LOCA, RO-C-AOP-D16 K/A 006000 2.4.9 Emergency Core Cooling System (ECCS)

Knowledge of low power / shutdown implications in accident (e.g loss of coolant accident or loss of residual heat removal) mitigation strategies.

RO - 3.8 SRO - 4.2 CFR - 41.10 / 43.5 / 45.13 K/A Match Candidate is required to demonstrate knowledge on correct procedure entry to mitigate a shutdown LOCA given a description of plant conditions.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO-C-AOP0250412-E3) explain the procedural mitigation strategy for a Mode 4 LOCA (RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 175 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 88 ID: NRCAUDIT07-0917 Points: 1.00 SRO Only Given the following conditions:

• The Plant is at 100% power.

• Reactor trip breaker testing is being performed with Reactor Trip Bypass Breaker B (52/BYB) racked in and closed.

• Both Reactor Trip Breakers are closed.

• Reactor Trip Bypass Breaker A is open and racked out.

What would be the consequences and required actions if the Train A Output Bay Mode Selector Switch was placed to TEST instead of the Train B switch?

A. A General Warning on Train B only. Reactor would NOT trip. Enter TS 3.0.3 due to 2 trains of the Reactor Protection System being inoperable.

B. A General Warning on Train A only. Reactor would NOT trip. Initiate a Manual reactor trip and enter OHP-4023-E-0, Reactor Trip or Safety Injection, since 2 trains of the Reactor Protection System are inoperable.

C. A General Warning on both RPS trains causing all Reactor Trip and Bypass Breakers to receive a trip signal. Enter OHP-4023-E-0, Reactor Trip or Safety Injection, to stabilize the plant.

D. A General Warning on Train B only which would result in opening the Reactor Trip A and Bypass B breakers only. Enter OHP-4023-E-0, Reactor Trip or Safety Injection, to stabilize the plant.

Answer: C Answer Explanation:

A. INCORRECT - With a General Warning on both trains the Reactor trips. Plausible if a general warning relay is actuated only for closure of bypass breaker B.

B. INCORRECT - With a General Warning on both trains the Reactor trips. Plausible if a general warning relay is actuated only for selecting Output Mode Selector switch to TEST.

C. CORRECT - Having the BYP B breaker closed is already providing a General Warning for TRAIN B (Open Contact K524 TRAIN B). Placing the Train A Output Mode Selector to Test will generate a Train A General Warning. Two General warnings will cause both trains to Trip.

D. INCORRECT - All Trip Breakers get an open signal. Plausible if the Train under test breakers will not open. (e.g. only RTA and BYB breakers open).

DC COOK OPS Page: 176 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0917 Level/Difficulty: H/3 Comments:

REFERENCE:

RO-C-01101 K/A 012000 A2.03 Reactor Protection System Ability to (a) predict the impacts of the following malfunctions or operations on the RPS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

INCORRECT channel bypassing RO - 3.4 SRO - 3.7 CFR - 41.5 / 43.5 / 45.3 / 45.13 K/A Match Candidate is required to demonstrate ability to predict the impact of a RPS malfunction and based on prediction select a procedure to mitigate the consequences of the malfunction.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO-C-01100-E1) Explain the purpose(s) and/or function(s) of the following:

a. Reactor Protection System (RPS)

b. Engineered Safeguards Features Actuation System (ESFAS)

c. ATWAS Mitigation System Actuation Circuit (AMSAC)

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 177 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 89 ID: NRCAUDIT07-0960 Points: 1.00 SRO Only Unit 1 is at 100% power. The following plant conditions exist:

• Both Supplemental DGs are out of service due to an electrical control problem

• The 1CD Emergency Diesel Generator (EDG) was declared INOPERABLE today (Monday) at 0600.

• Engineering can NOT rule out EDG common mode failure

• It is estimated that 1 CD DG will not be returned to Operable status for 7 days.

What action is required?

REFERENCE PROVIDED A. Perform an operability run on the 1AB EDG by 0600 tomorrow AND restore one Supplemental DG by 0600 Thursday.

B. The unit must be in at least HOT STANDBY by 1200 today.

C. Perform an operability run on the 1AB EDG by 0600 tomorrow AND restore both Supplemental DGs by 0600 Thursday.

D. Restore both Supplemental DGs by 1800 today OR perform an operability run on the 1AB EDG by 0600 tomorrow.

Answer: C Answer Explanation:

A. INCORRECT - Both SDGs are required. Plausible if student believes that one SDG can support a safeguards train.

B. INCORRECT - This is the action for Condition B.4.1 not being met. Plausible if student does not understand that either B.4.1 (Common mode) or B.4.2 (Run the other EDG within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) can be met to comply with this condition.

C. CORRECT - Tech Spec requires action B to be entered. Action B1 requires both Supplemental DGs to be operable (Action C requires them to be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if EDG is not restored). Since common mode failure cannot be ruled out, the AB EDG must be run within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. INCORRECT - The SDGs must be restored within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> AND the AB EDG must be run.

Plausible if student does not completely read 3.8.1 B conditions to determine that SDGs and other EDG must both be addressed.

DC COOK OPS Page: 178 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0960 Comments:

REFERENCE:

Tech Spec 3.8.1 Reference Provided: TS/SR 3.8.1 K/A 064000 2.1.32 Emergency Diesel Generator (ED/G) System Ability to explain and apply system limits and precautions RO - 3.8 SRO - 4.0 CFR - 41.10 / 43.2 / 45.12 K/A Match Candidate is required to demonstrate knowledge of the system limits (Tech Spec operability) and the ability to determine the required actions based on the limits.

SRO - 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). [55.43(b)(2)]

Associated objective(s):

(RO-C-03200-E20) Given a description of plant conditions and/or the results of a Surveillance Test, determine the following: All applicable TS, System Operability, and most limiting LCO.

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 179 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 90 ID: 2008NRC-0605-SRO Points: 1.00 SRO Only 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?

REFERENCE PROVIDED 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Containment Purge/Pressure Relief operations may continue for up to 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 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: 180 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: 2008NRC-0605-SRO External Topic ID: TS01-3: Technical Specifications/TRM - Application1350: Radiation Monitoring System Level/Difficulty: H/3 Comments:

REFERENCE:

Tech Spec 3.3.6, TRM 8.3.8 Reference Provided - TS 3.3.6, TRM 8.3.8, and the associated Bases.

K/A 073000 A2.02 Process Radiation Monitoring (PRM) System Ability to (a) predict the impacts of the following malfunctions or operations on the PRM System and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations:

Detector failure RO - 2.7 SRO - 3.2 CFR - 41.5 / 43.5 / 45.3 / 45.13 K/A Match Candidate is required to demonstrate ability to identify the impacts of a PRM failure and apply Tech Spec restrictions to control operation of the Containment purge System.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO-C-01350-E9) Given a description of plant condition and/or the results of a Surveillance Test, determine the following: All applicable TS, System Operability, and most limiting LCO.

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 181 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 91 ID: RO26-0176A-SRO Points: 1.00 SRO Only Given the following conditions on Unit 2:

• Unit was at 90% power.

• Control Rods were in AUTO.

• CBD began to step out with no mismatch signal.

• Rods were taken to MANUAL and rod motion ceased.

The following conditions now exist:

• CBD Bank Demand position is now 222 Steps.

• Group 2 Bank D Rod Position Indicators (RPIs) ALL indicate 222 Steps.

• Group 1 Bank D RPIs indicate as follows:

• Rod D4: 205 Steps.

• Rod D12: 222 Steps.

• Rod M12: 207 Steps.

• Rod M4: 222 Steps.

Reactor Engineering has determined that all CBD rods are free to move and has provided the following information:

• R is 1.041

• CFQ is 2.335

• K(Z) is .95 (at 10 feet)

• FwQ(Z)_ is 2.174 (at 10 feet)

Which ONE of the following identifies the Technical Specification Action Condition(s) that must be entered?

REFERENCE PROVIDED A. 3.1.4.A Only B. 3.1.4.B Only C. 3.1.4.A and 3.1.4.B Only D. 3.1.4.B and 3.1.4.D Only Answer: D Answer Explanation:

A. INCORRECT - Rods are free to move, as stated in the stem, thus they are all operable.

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 B. INCORRECT - Using the numbers provided with Figure 3.1.4-1, rods which are <14 steps misaligned meet alignment requirements. This is the correct action with only 1 misaligned rod.

C. INCORRECT - Rods are free to move and are operable.

D. CORRECT - Both D4 and M12 are unacceptably misaligned. D4 is 17 steps off and M12 is 15 steps off.

Question ID: RO26-0176A-SRO External Topic ID: TS01-3: Technical Specifications/TRM - Application Level/Difficulty: H/3 Comments:

REFERENCE:

Tech Spec 3.1.4 Reference Provided: Tech Spec 3.1.4 Rod Group Alignment Limits K/A 014000 A2.04 Rod Position Indicating System (RPIS)

Ability to (a) predict the impacts of the following malfunctions or operations on the RPIS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Misaligned rod.

RO - 3.4 SRO - 3.9 CFR - 41.5 / 43.5 / 45.3 / 45.13 K/A Match Candidate is required to demonstrate ability to recognize whether the misaligned rod is Operable or not and based on that decision determine which tech Spec entered to control /

mitigate the malfunctions.

SRO - 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). [55.43(b)(2)]

Associated objective(s):

(RO-C-01200-E24) EXPLAIN the basis for Technical Specification LCOs, Action Statements, and Surveillance Requirements associated with the Rod Position Indication System.

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 183 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 92 ID: RO-C-01301-E12-2 Points: 1.00 SRO Only Due to a wire connector issue, it was determined that a majority of the Core Exit Thermocouples (CET's) were Inoperable.

A review of the CET's that are available in each quadrant are listed below Quadrant Number of Thermocouples Operable 1 4 2 1 3 0 4 3 Which one of the following describes the required Technical Specification actions if any?

REFERENCE PROVIDED A. No Technical Specification actions are required for the CET's since there are 2 channels operable.

B. Restore 1 CET in Quadrant 3 within 7 days and an additional CET's in both Quadrant 2 and 3 within 30 days.

C. Immediately restore a CET in Both Quadrants 2 and 3 or be in mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

D. Immediately initiate action in accordance with Specification 5.6.6 and be in mode 3 in 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and Mode 4 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Answer: B Answer Explanation:

A. INCORRECT - Tech Specs require 2 CETs (channels) in each Quadrant (function) Quadrant 2 only has 1 and 3 has none.

B. Correct - Quadrant with 1 channel inoperable requires action A (30days), Quadrant 3 requires Actions A (30 days for both) and D (7 days for at least 1)

C. INCORRECT - Action F is only required if the actions associated with D are not met. It would only be required for quadrant 3.

D. INCORRECT - Action G is not required for these functions. Action F is only required if the actions associated with D are not met.

DC COOK OPS Page: 184 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO-C-01301-E12-2 Level/Difficulty: H/2 Comments:

REFERENCE:

Tech Spec 3.3.3 Reference Provided: Tech Spec 3.3.3 PAM Instrumentation K/A 017000 2.2.40 In-Core Temperature Monitor (ITM) System 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 K/A Match Candidate is required to demonstrate ability to determine the impact of the CET failures (fails low) and then use Tech Specs to determine the correct limits and actions for restoration.

SRO - 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). [55.43(b)(2)]

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 (RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 185 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 93 ID: NRCAUDIT07-0083 Points: 1.00 SRO Only Unit 2 is in Mode 6 - Refueling with Fuel Movement in progress.

The following conditions exist at 0100:

• Source range channels N31 and N32 are operable.

• Source range channel N31 is selected for audible function.

• Gamma-Metrics channels N21 and N23 are operable.

Given the following events and times:

• At 0200, Source range channel N32 fails.

• At 0300, Source range channel N31 fails.

• At 0400, Gamma-Metrics channel N21 fails.

• At 0500, Gamma-Metrics channel N23 fails.

Which ONE of the following is the earliest time that core alterations must be suspended?

A. 0200 B. 0300 C. 0400 D. 0500 Answer: B Answer Explanation:

A. INCORRECT - Wide range flux monitors (gamma metrics) are allowed to substitute for NIS source flux monitors. 2 WR and 1 SR with Audible are operable.

B. CORRECT - Tech Spec 3.9.2 requires 2 operable source range channels. Wide range flux monitors (gamma metrics) are allowed to substitute for NIS source flux monitors. BUT the gamma metrics do NOT provide an audible function.

C, INCORRECT - Gamma metrics do NOT provide an audible function.

D. INCORRECT - 2 Channels and an Audible are required.

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: NRCAUDIT07-0083 Comments:

REFERENCE:

Tech Spec 3.9.2; 02-OHP-4030-227-037, Refueling Surveillance, Data Sheet 3, Item 2 K/A 034000 A4.02 Fuel Handling Equipment System (FHES)

Ability to manually operate and/or monitor in the control room:

Neutron levels RO-3.5 SRO-3.9 CFR - 41.7 / 45.5 to 45.8 K/A Match Candidate is required to demonstrate ability to monitor Fuel handling System availability based on Audible Count rate functioning in Control room.

SRO - 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). [55.43(b)(2)]

Associated objective(s):

(RO-C-ADM13-E3) Given a set of plant conditions determine if refueling operations can be initiated or continue.

(RO31 NRC Exam 2018 76-100 SRO)

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 94 ID: NRCAUDIT07-0064 Points: 1.00 SRO Only Both Units are at full power with three hours left until shift turnover with the following shift manning:

Position Unit-1 Unit-2 Shared Shift Manager 1 WCC-SRO 1 Unit Supervisor 1 1 Reactor Operator 2 2 Qualified Operator (AEO) 3 3 2 Shift Technical Advisor 1 The Unit 1 Unit Supervisor and a Unit 1 Qualified Operator are injured during a mid-shift plant tour and are taken to first aid for treatment and observation.

If this condition exists until shift turnover, per Tech Specs, Unit 1 minimum shift crew composition is:

A. met provided the STA assumes the Unit 1 Unit Supervisor position.

B. met provided the WCC-SRO assumes the Unit 1 Unit Supervisor position.

C. NOT met and the unit must enter T.S.3.0.3 if the proper crew composition is not met within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

D. NOT met and no further action is required other than the continual efforts to restore crew composition.

Answer: B Answer Explanation:

A. INCORRECT - Plausible if the STA is licensed, but Shift Technical Advisor is still a separate required position.

B. Correct - Per Technical Specifications 1 Unit Supervisor per unit is required. The WCC-SRO is not a Tech Spec required position so he may assume the Unit Supervisor position.

C. INCORRECT - Plausible since a T.S./TRM requirement is not met, but T.S. 3.0.3 is Not Applicable for requirements of T.S. Section 5 DC COOK OPS Page: 188 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 D. INCORRECT - Plausible since the U.S. can be vacant for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> while efforts are taken to fill the position, however there are 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> remaining until turnover; the Unit 1 Supervisor position needs to be filled.

Question ID: NRCAUDIT07-0064 Comments:

REFERENCE:

Tech Spec 5.2.2, 10CFR50.54 K/A 194001 2.1.4 Generic Conduct of Operations Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operations, maintenance of active license status, 10CFR55, etc.

RO-3.3 SRO-3.8 CFR - 41.10 / 43.2 K/A Match Candidate is required to demonstrate knowledge of licensed operator responsibilities related to shift staffing minimum requirements.

SRO - Facility operating limitations in Section 5.2.2 of TS concerning shift staffing. [55.43(b)(2)]

Associated objective(s):

(RO-C-ADM14-E18) Describe the requirements for Shift Staffing as given in OHI-4000 Attachment 23,Shift Staffing" (RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 189 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 95 ID: CM-7201 Points: 1.00 SRO Only The plant is in MODE 1.

The purpose of the RCS Specific Activity Technical Specification is to maintain analysis assumptions with regard to the __________.

A. Stuck Open Pressurizer Pressure Operated Relief Valve (PORV) Event.

B. Boron Dilution Event.

C. Steam Generator Tube Rupture (SGTR) Accident.

D. Rod Ejection Accident.

Answer: C Answer Explanation:

A. INCORRECT - Plausible as Stuck Open PORV would result in PRT rupture and release of activity but the release would be inside containment.

B. INCORRECT - Plausible if candidate confused activity with reactivity concern.

C. CORRECT - RCS Specific Activity limit is based on limiting the exposure to the general public at the site boundary during releases occurring during a Steam Generator Tube Rupture.

D. INCORRECT - Plausible as Rod ejection accident would result in a LOCA and a release of activity but the release would be inside containment.

Question ID: CM-7201 External Topic ID: TS01-4: Technical Specifications/TRM - Bases Comments:

REFERENCE:

Tech Spec 3.4.16 Bases; Applicable Safety Analysis Section DC COOK OPS Page: 190 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 194001 2.2.5 Conduct of Operations Knowledge of primary and secondary plant chemistry limits.

K/A Match Candidate is required to demonstrate knowledge on bases for primary chemistry limit for RCS Specific activity.

SRO - Facility operating limitations in the TS and their bases. Knowledge on bases for primary chemistry limit [55.43(b)(2)]

Associated objective(s):

(RO-C-TS01-E11) Given 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 (RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 191 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 96 ID: RO26-0150A-SRO Points: 1.00 SRO Only When a modification, test, or experiment is proposed which may increase the probability of occurrence or consequences of an accident, the activity must be scrutinized using a multi-phase process.

Which part of the process CONCLUDES whether the Station must obtain NRC approval PRIOR to carrying out the modification, test or experiment?

A. Applicability Determination B. 10CFR50.59 Screening C. 10CFR50.59 Evaluation D. Licensing Amendment Answer: C Answer Explanation:

A. INCORRECT - The Applicability Determination determines if 50.59 is applicable.

B. INCORRECT - The screening determines if an evaluation must be done.

C. CORRECT - The evaluation process is where the conclusion is made regarding whether prior approval is required.

D. INCORRECT - A license amendment may be required based on the outcome of this process, but the amendment is not part of the 10CFR50.59 evaluation.

Question ID: RO26-0150A-SRO External Topic ID: ADMN-03: Work Control Process Level/Difficulty: F/3 Comments:

REFERENCE:

PMP-2350-SES-001 DC COOK OPS Page: 192 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 194001 2.2.5 Equipment Control Knowledge of the process for making design or operating changes to the facility.

RO - 2.2 SRO - 3.2 CFR - 41.10 / 43.3 / 45.13 K/A Match Candidate is required to demonstrate knowledge of where in the process the conclusion is made "if the change affects the licensing basis."

SRO - Knowledge of facility Licensee Procedures Required to Obtain Authority for Design and Operating Changes in the Facility. Screening and evaluation process under 10CFR 50.59 Changes, Tests and Experiments.[55.43(b)(3)]

Associated objective(s):

(RO-C-ADM03-E13) Given a set of conditions determine whether or not a 50.59 safety screening is required.

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 193 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 97 ID: 2008NRC-0631 Points: 1.00 SRO Only Given the following on Unit 2:

• You are the Unit Supervisor

• The Turbine driven AFW is INOPERABLE due to a failed trip/throttle valve.

• It was discovered that the West Motor Driven AFW Pumps bearing has failed rendering it INOPERABLE.

• TS 3.7.5 Auxiliary Feedwater was entered due to 2 AFW trains INOPERABLE.

Power was being reduced to 15% in preparation for a Plant Shutdown as required per TS 3.7.5 Auxiliary Feedwater (AFW) System when the reactor was tripped.

The plant has been stabilized with the East Motor Driven AFW pump operating and the crew has completed 02-OHP-4023-ES-0.1.

The STA states that you should stabilize the plant at 375°F and NOT enter Mode 4.

Do you agree or disagree and why?

A. Agree, Tech Spec 3.0.4 prohibits mode changes if all applicable tech specs for that mode are not met.

B. Agree, with 2 trains of AFW INOPERABLE the plant should be maintained in the current mode until at least 2 AFW pumps are OPERABLE.

C. Disagree, a standing Notice of Enforcement Discretion is in place to allow the plants to continue to lower modes even if they don't meet all Tech Specs.

D. Disagree, the plant should be placed into Mode 4 where only 1 AFW pump is required to be OPERABLE.

Answer: D Answer Explanation:

A. INCORRECT - Tech Spec 3.0.4 allows you to pass through lower modes without meeting all conditions ONLY if you are complying with a required action statement of another Tech Spec. Plausible since TS does not allow you to enter higher modes.

B. INCORRECT - The TS 3.7.5 contains a note that 3.0.3 and all LCOs requiring Mode changes are suspended if ALL 3 AFW trains are Inoperable. There are only 2 trains Inoperable.

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 C. INCORRECT - A standing Notice of Enforcement Discretion does NOT exist. Plausible since a NED may be issued as a short term solution for failing to meet TS with adequate justification.

D. CORRECT - Tech Spec 3.0.4 allows you to pass through lower modes without meeting all conditions ONLY if you are complying with a required action statement of another Tech Spec. Plausible since TS does not allow you to enter higher modes.

Question ID: 2008NRC-0631 Comments:

REFERENCE:

Tech Spec 3.7.5 & 3.0.4 and Bases B 3.0.4 K/A 194001 2.2.38 Equipment Control Knowledge of conditions and limitations in the facility license.

RO - 3.6 SRO - 4.5 CFR - 41.7 / 43.1 / 45.13 K/A Match Candidate is required to demonstrate knowledge of the license conditions relating to mode changes and Tech Specs.

SRO - 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 AFW requirements. [55.43(b)(2)]

Associated objective(s):

(RO-C-ADM04-E1) Describe the Technical Specification requirements for testing, including the types of things that are required to be tested.

(RO31 NRC Exam 2018 76-100 SRO)

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EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 98 ID: RO26-0026-SRO Points: 1.00 SRO Only Given the following plant conditions:

• Unit 2 is in Mode 5

• Actions are being taken to start the Containment Purge system in the Clean-Up Mode.

• Due to an inadvertent overload trip of the Containment Purge supply fan, there has been a 30 hour3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br /> delay since Radiation Protection (RP) approval to start the purge.

An operator is ready to start aligning purge and asked for Unit Supervisor permission to continue.

Which ONE of the following describes the appropriate response by the Unit Supervisor?

A. Purge may be initiated since RP approval was within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of the start of the purge operation.

B. Purge may be initiated as long as Plant Manager approval is obtained within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of purge initiation.

C. Purge can NOT be placed in service due to exceeding 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time of initial RP approval.

D. Purge can NOT be placed in service until a second air sample is obtained from the containment atmosphere.

Answer: C Answer Explanation:

A. INCORRECT - Purge must be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of RP approval.

B. INCORRECT - Plant Manager approval is required prior to purge operation for purges that are not approved by RP.

C. CORRECT - Purge must be initiated within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of RP approval.

D. INCORRECT - RP approval is required within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of the purge initiation.

DC COOK OPS Page: 196 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 Question ID: RO26-0026-SRO External Topic ID: 2800: Containment Ventilation Level/Difficulty: H/3 Comments:

REFERENCE:

2-OHP-4021-028-005, Attachment 1, Step 3.1.

K/A 194001 2.3.6 Radiation Control Ability to approve release permits.

RO - 2.0 SRO - 3.8 CFR - 41.13 / 43.4 / 45.10 K/A Match Candidate is required to demonstrate ability to determine if the requirements for performing a containment purge (radiation release) have been satisfied.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO-C-02800-E12) Explain the basis for the following Containment Ventilation System Technical Specification LCOs, Action Statements, and Surveillance Requirements: {Generic 2.1.10 - 2.7/3.9}

a. 3.6.3, Containment Isolation Valves

b. 3.6.4, Containment Pressure

c. 3.6.5, Containment Air Temperature

d. 3.6.10, Containment Air Recirculation/Hydrogen Skimmer (CEQ) System (RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 197 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 99 ID: CM-1027 Points: 1.00 SRO Only While recovery of a SGTR is in progress, the PET informs the control room to cooldown using ES-3.1, Post SGTR Cooldown Using Backfill. In Step 2 of ES-3.1, the Pressurizer heaters are placed in service to saturate the PZR.

Which ONE of the following statements describes the reason for this step?

A. Raises RCS pressure above ruptured steam generator pressure so RCS leakage is into the steam generator to help cool it down.

B. Allows mixing in the PRZ, thus preventing stratification of boron to prevent a loss of SDM later in cooldown.

C. Minimizes RCS pressure transients on the opening of a PRZ spray valve or PORV later in the procedure.

D. Raises RCS pressure so RCPs can be started for cooldown if they were secured previously.

Answer: C Answer Explanation:

A. INCORRECT - Plausible if student believes that RCS pressure is controlled above ruptured SG pressure to refill ruptured SG at low level.

B. INCORRECT - Plausible if student believes that RCS boration to CSD SDM requirements requires the PRZ to be equally mixed with RCS.

C. CORRECT - Pressure reduction in the vapor space will be buffered by PRZ liquid flashing to steam if the liquid is at saturation conditions.

D. INCORRECT - Plausible if student does not understand that requirement for PRZ liquid to be saturated at ruptured SG pressure before starting a RCP predicts better pressure control when PRZ level drops to fill head void after pump start.

Question ID: CM-1027 External Topic ID: PROC-3: Procedure Knowledge Level/Difficulty: 3.1/3.2 Comments:

REFERENCE:

02-OHP-4023-ES-3.1 DC COOK OPS Page: 198 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 194001 2.4.18 Emergency Procedures / Plan Knowledge of the specific bases for EOPs RO - 3.3 SRO - 4.0 CFR - 41.10 / 43.1 / 45.13 K/A Match Candidate is required to demonstrate knowledge of the bases for re-energizing the pressurizer heaters during a SGTR.

SRO - 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. [55.43(b)(5)]

Associated objective(s):

(RO-C-EOP08-E18) For the E-3 series procedures and the ECA-3 series procedures discuss the basis or reason for all Steps.

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 199 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 100 ID: CM-1058 Points: 1.00 SRO Only Which ONE of the following sets of duties MAY be delegated by the Site Emergency Coordinator?

1) Directing the notification of plant personnel.

2) Performing mitigating actions as required by the EOPs.

3) Classification of the emergency.

4) Performing initial off-site dose assessment.

5) Approving protective action recommendations.

6) Directing the notification of off-site officials.

A. 1, 2, 4 B. 1, 3, 5 C. 2, 4, 6 D. 3, 5, 6 Answer: A Answer Explanation:

A. CORRECT - Notification of PLANT personnel, Direction of Mitigating actions, and Performance of Dose Assessment activities May Be performed by others.

B. INCORRECT - Classification of the event and PAR approval may NOT Be Delegated.

C. INCORRECT - Directing Notification of Off-site officials may NOT Be Delegated.

D. INCORRECT - These are the Duties that may NOT Be Delegated.

Question ID: CM-1058 External Topic ID: EP03-1: Emergency Plan - Event Classification Level/Difficulty: F/3 Comments:

REFERENCE:

PMP-2080-EPP-100, Emergency Response Sec 3.1.3 DC COOK OPS Page: 200 of 201 26 February 2018

EXAMINATION ANSWER KEY RO31 NRC EXAM 2018 K/A 194001 2.4.29 Emergency Procedures/Plan Knowledge of the emergency plan.

RO - 3.1 SRO - 4.4 CFR - 41.10 / 43.5 / 45.11 K/A Match Candidate is required to demonstrate knowledge of the SRO duties which actions/activities that be performed by the emergency coordinator vs. those that can be delegated.

SRO - Question has candidate demonstrate knowledge of administrative procedure requirements of duties which can be delegated by the SM / SEC during Emergency plan implementation. [55.43(b)(5)]

Associated objective(s):

(RO31 NRC Exam 2018 76-100 SRO)

DC COOK OPS Page: 201 of 201 26 February 2018