Text

CG-2021-02-DRAFT Written Exam
	ML21113A005
Person / Time
Site:	Columbia
Issue date:	03/04/2021
From:	Greg Werner; Operations Branch IV
To:	; Energy Northwest
References
	50-397/21-02 50-397/OL-02
	Download: ML21113A005 (574)
	v • d • e

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1 Examination Outline Cross-reference: 1 Revision: 0 Date: 10/29/20 Tier: 1 Group: 1 K/A Number: 295001.AK2.04 Level of Difficulty: 3 RO Importance Rating: 3.3 K/A

Description:

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the following: Reactor/turbine pressure regulating system CGS is in Mode 1.

A loss of both RRC pumps occurs.

The crew performs a manual reactor scram.

The CRS enters PPM 5.1.1 - RPV Control.

Once the plant is stabilized, the CRS enters ABN-RRC-LOSS, Loss of Reactor Recirculation Flow.

The CRS directs maintaining RPV pressure 600-1000 psig.

MSIVs are open.

How should RPV pressure be maintained?

Maintain RPV pressure using...

A. DEH in automatic at no more than 8 psig/min.

B. SRVs at no more than 50 psig/min.

C. DEH in manual, disregard cooldown rate.

D. BPVs in manual, in 200 psig increments.

Answer: D K/A Match:

Requires an understanding of how the RPV/Turbine Pressure Regulating system is operated on a loss of both RRC pumps.

Explanation:

A. Incorrect. Plausible since this is a method of pressure control used in PPM 5.1.1. However, ABN-RRC-LOSS directs controlling RPV pressure in 200 psig batches using SRVs or BPVs to reduce thermal stratification.

B. Incorrect. Plausible since this is a method of pressure control used in PPM 5.1.1. However, ABN-RRC-LOSS directs controlling RPV pressure in 200 psig batches using SRVs or BPVs to reduce thermal stratification.

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1 C. Incorrect. Plausible since depressurizing using DEH at the maximum rate would help reduce stratification. However, ABN-RRC-LOSS directs controlling RPV pressure in 200 psig batches using SRVs or BPVs to reduce thermal stratification.

D. Correct. ABN-RRC-LOSS, step 4.1.2.b, directs controlling RPV pressure in 200 psig batches using SRVs or BPVs to reduce thermal stratification Tier 1 Discussion Requires knowledge of abnormal procedure supplemental actions.

Technical Reference(s)

ABN-RRC-LOSS, Loss of Recirculation Flow Attached w/ Revision #

PPM 5.1.1, RPV Control See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5023 - Predict the effect that a loss or malfunction of the following will have on the Reactor Recirculation System: c. Dual RRC Pump trip Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Need to synthesize the initial conditions given in the question stem with the requirements of PPM 5.1.1 and ABN-RRC-LOSS 10 CFR Part 55 Content: 55.41 10

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1 Comments /

Reference:

ABN-RRC-LOSS Rev: Major: 16 Minor: N/A

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-1 Comments /

Reference:

PPM 5.1.1 Rev: Major: 22 Minor: N/A

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-2 Examination Outline Cross-reference: 2 Revision: 0 Date: 10/14/20 Tier: 1 Group: 1 K/A Number: 295003.AK2.04 Level of Difficulty: 3 RO Importance Rating: 3.4 K/A

Description:

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C. POWER and the following: A.C. electrical loads CGS is in Mode 1.

Transformer TR-B is out of service.

A breaker fault causes breaker E-CB-7/1 to open.

DG-1 starts and repowers SM-7.

Operators are performing actions in accordance with ABN-ELEC-SM1/SM7, Distribution System Failures.

How should the Steam Tunnel Fans be operated?

A. RRA-FN-8 should be operating. Restart RRA-FN-9 from the control room.

B. RRA-FN-9 should be operating. Start RRA-FN-21 from the control room.

C. Both RRA-FN-8 and RRA-FN-9 should be operating. No further actions are required.

D. Both RRA-FN-8 and RRA-FN-9 should be operating. Start RRA-FN-21 from the control room and run 3 fans until Steam Tunnel temperatures return to normal.

Answer: B K/A Match:

Requires knowledge of condition of AC electrical loads following a loss of A.C. power to a bus and the actions required by the ABN.

Explanation:

A. Incorrect. Plausible if believed that RRA-FN-8 is powered by SM-8. However, RRA-FN-8 is powered from SM-7 via SL-71. When E-CB-7/1 opens, SM-7 loses power until it is repowered from DG-1. RRA-FN-8 trips and does not automatically restart when power is restored. It must be restarted from the field. Additionally, RRA-FN-9 will be running and cannot be started from the control room.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-2 B. Correct. When E-CB-7/1 opens, SM-7 loses power until it is repowered from DG-1. RRA-FN-8 trips and does not automatically restart when power is restored. That leaves RRA-FN-9 as the only Steam Tunnel Fan operating. ABN-ELEC-SM1/SM7, step 4.2.1, states that operators must verify two steam tunnel fans are operating. The operator should start Backup Steam Tunnel Fan RRA-FN-21 from the control room. Note that it would also be correct to restart RRA-FN-8 from the field once DG-1 has repowered SM-7. However, this is not an answer choice.

C. Incorrect. Plausible if it is believed that RRA-FN-8 will automatically restart when SM-7 is repowered from DG-1. However, RRA-FN-8 does not automatically restart and must be restarted from the field. RRA-FN-9 is the only steam tunnel fan in operation. Since ABN-ELEC-SM1/SM7, step 4.2.1 requires operators to verify that two steam tunnel fans are operating, a second fan must be started.

D. Incorrect. Plausible if it is believed that RRA-FN-8 will automatically restart when SM-7 is repowered from DG-1. However, RRA-FN-8 does not automatically restart and must be restarted from the field. Additionally, the procedure caution in ABN-ELEC-SM1/SM7, step 4.2.1 prohibits running 3 fans concurrently in Mode 1.

Tier 1 Discussion Meets requirements for Tier 1 questions since examinees are required to demonstrate knowledge of Abnormal Procedure steps and cautions.

Technical Reference(s)

ABN-ELEC-SM-7, SM-1, SM-7, SM-75, SM-72, SL-71, SL-73 & SL-11 Distribution System Failures Attached w/ Revision #

SD000183, RB HVAC See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: None Learning Objective: 15755 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-ELEC-SM1/SM7.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires candidate to synthesize information given with a knowledge of actions and cautions of the abnormal procedure along with an understanding of the power supplies to steam tunnel fans along with equipment response to a momentary loss of power Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-2 10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

ABN-ELEC-SM1/SM7 Rev: Major: 022 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-2 Comments /

Reference:

SD000183 Rev: Major: 12 Minor: 2 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-2 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3 Examination Outline Cross-reference: 3 Revision: 0 Date: 11/23/20 Tier: 1 Group: 1 K/A Number: 295004.AK3.03 Level of Difficulty: 2 RO Importance Rating: 3.1 K/A

Description:

Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER : Reactor SCRAM CGS is in Mode 1.

Reactor power is 100%.

A fault has occurred on DC bus S1-7.

S1-7 voltage as indicated and lowering at 10 VDC per minute.

What action should be taken?

A. Scram the reactor due to imminent loss of both Reactor Recirculation Pumps.

B. Manually transfer SM-1 and SM-3 to TR-S since automatic transfer capability will be lost.

C. Scram the reactor and trip the main turbine to automatically transfer to TR-S before this capability is lost.

Page 1 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3 D. Manually transfer inverter IN-5 to the Alternate AC source since automatic transfer capability will be lost.

Answer: C K/A Match:

Requires understanding of the reason for performing a manual reactor scram when a loss of BOP DC is imminent.

Explanation:

A. Incorrect. Plausible since a reactor scram should be initiated and SH-5 and SH-6 supply breakers lose control power on a loss of S1-7. However, the buses will not be lost and therefore, RRC pumps will remain operating.

B. Incorrect. Plausible since automatic transfer from TR-S to TR-N is lost on a loss of S1-7.

However, automatic transfer of SM-7 and SM-8 to TR-B will still occur and the procedure simplifies the operation of the AC distribution system by requiring an automatic scram and main turbine / main generator trip to ensure that an automatic transfer to TR-S occurs prior to S1-7 voltage going below 105 VDC.

C. Correct. In accordance with ABN-ELEC-125VDC, Plant BOP, DIV1,2, &3 125 VDC Distribution System Failures, a loss of S1-7 (BOP 125 VDC) will cause a loss of breaker control power and the loss of ability to manually or automatically operate breakers remotely. In this condition, a MG trip might not cause an automatic transfer from TR-N to TR-S. A CAUTION prior to the immediate action step details that a loss of control power may occur at DC voltages less than 105 VDC. The indications given in the question stem show that voltage will reach this level in approximately 30 seconds. Therefore, the crew should perform step 4.1.1 to ensure that AC buses automatically transfer to TR-S.

D. Incorrect. Plausible since S1-7 supplies inverter IN-5. However, the normal power supply to IN-5 is MC-8A. Normal operation of IN-5 will continue as long as MC-8A is energized.

Tier 1 Discussion Requires knowledge of abnormal procedure cautions and required actions.

Technical Reference(s)

ABN-ELEC-125VDC, Plant BOP, DIV1,n2, &3 125 VDC Distribution System Failures Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7654 - Predict the effect(s) a failure of 125VDC bus S1-7 will have on:

Page 2 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3

a. IN5

b. RFW

c. Main Turbine

d. SM1,2,3 Control Power

e. SH5,6 Control Power

f. Generator Breaker Control Power

g. CR Annunciators Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information in the question stem with a knowledge of conditions where S1-7 is considered lost along with an understanding of automatic actions that will occur with a loss of S1-7 and actions that are required to be completed.

10 CFR Part 55 Content: 55.41 10 Page 3 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3 Comments /

Reference:

ABN-ELEC-125VDC Rev: Major: 16 Minor: N/A Page 4 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3 Page 5 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-3 Page 6 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Examination Outline Cross-reference: 4 Revision: 1 Date: 10/22/20 Tier: 1 Group: 1 K/A Number: 295005.AK1.01 Level of Difficulty: 2 RO Importance Rating: 4.0 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP: Pressure effects on reactor power.

CGS is in Mode 1.

The reactor has been operating at full power for 335 days since the last refueling outage.

A Main Turbine trip occurs, followed by an automatic reactor scram.

The CRS enters PPM 3.3.1, Reactor Scram, and PPM 5.1.1, RPV Control.

CRO1 is checking status of Reactor Recirculation (RRC) pumps in accordance with the subsequent actions of PPM 3.3.1.

What is the expected RRC pump response to this event and what is the reason for this response?

The RRC pumps are (1) to minimize the effect of (2) .

A. (1) running at 15 Hz (2) low RPV water level B. (1) running at 15 Hz (2) reactor power increase due to rising pressure C. (1) tripped (2) low RPV water level D. (1) tripped (2) reactor power increase due to rising pressure Answer: D Page 1 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 K/A Match:

Requires knowledge of the operational implications s of a main turbine trip on reactor pressure and reactor power.

Explanation:

A. Incorrect. Plausible since a reactor scram will cause a RRC pump runback to 15 Hz due to low RPV water level. However, with the conditions listed in the question stem, RRC pumps will trip when the Main Turbine trips due to the EOC-RPT circuit.

B. Incorrect. Plausible since RRC pump response is based on mitigating a power increase due to an increase in pressure. However, for the conditions given, RRC pumps will trip.

C. Incorrect. Plausible since RRC pumps will trip for the conditions given. However, the reason for this response is not due to low RPV level. It is based on mitigating a power increase due to an increase in pressure.

D. Correct. When the Main Turbine trips, the load reject will cause steam flow to rapidly lower. This will cause reactor pressure to initially rise. The higher reactor pressure will cause reduced voids in the core, adding positive reactivity and causing reactor power to rise. This effect is more pronounced at EOC where rod worth is lower due to fuel depletion. To counteract this effect, RRC pumps automatically trip on a scram with initial power GE 29.5% at EOC.

Tier 1 Discussion This question requires knowledge of expected plant conditions when performing subsequent actions of emergency operating procedures. Therefore, it meets the requirements for a Tier 1 question.

Technical Reference(s)

SD000178, Reactor Recirculation System Description Attached w/ Revision #

SD000184, Reactor Recirculation Flow Control (ASD)

See Comments / Reference PPM 3.3.1, Reactor Scram Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5023e - Predict the impact on the RRC System of each of the following conditions or events: e. EOC-RPT logic Question Source: #: Bank #

LO01783 (Note changes or attach parent)

Question History: Last NRC Exam: 2009 Question Cognitive Level:

Page 2 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Justification for Cognitive Level Examinee must evaluate question conditions using an understanding of the relationship between reactor pressure and reactor power along with a knowledge of the RRC pump response at EOC, along with expected plant conditions while performing emergency operating procedures.

10 CFR Part 55 Content: 55.41 1 Comments /

Reference:

SD000178 Rev: Major: 17 Minor: 1 Page 3 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Comments /

Reference:

SD000184 Rev: Major: 20 Minor: 2 Page 4 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Comments /

Reference:

SD000184 Rev: Major: 20 Minor: 2 Page 5 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-4 Comments /

Reference:

Parent Question LO01783 Rev: Major: NA Minor: NA Page 6 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-5 Examination Outline Cross-reference: 5 Revision: 0 Date: 10/14/20 Tier: 1 Group: 1 K/A Number: 295006.AA1.07 Level of Difficulty: 2 RO Importance Rating: 4.1 K/A

Description:

Ability to operate and/or monitor the following as they apply to SCRAM : Control rod position CGS is in Mode 1.

The crew manually scrams the reactor due to an instrument failure.

The CRS enters PPM 5.1.1, RPV Control.

Operators are performing the following step:

Considering control rod position alone, which of the following combinations of control rod positions will provide sufficient shutdown margin to assure that the reactor is shutdown under all conditions?

One control rod at position (1) , an additional control rod at position (2) , and the remaining control rods at position (3) .

A. (1) 08 (2) 04 (3) 02 B. (1) 48 (2) 02 (3) 04 C. (1) 08 (2) 04 (3) 00 D. (1) 48 (2) 02 (3) 00

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-5 Answer: D K/A Match:

Requires knowledge of required rod position for adequate shutdown margin following a scram.

Explanation:

A. Incorrect. Plausible if it is believed that 2 control rods may be greater than position 02 and ensure adequate shutdown margin. However, only 1 control rod may be greater than position 02.

B. Incorrect. Plausible if it is believed that all control rods may be at position 04 (with one greater than position 04) and still demonstrate adequate shutdown margin. However, all control rods must be no greater than position 02 with one greater than position 02.

C. Incorrect. Plausible if it is believed that all control rods may be at position 04 (with one greater than position 04) and still demonstrate adequate shutdown margin. However, all control rods must be no greater than position 02 with one greater than position 02.

D. Correct. In accordance with OI-15, EOP and EAL Clarifications, The reactor is shutdown under all conditions if no more than one controlis withdrawn past position 02. Any number of rods may be partially withdrawn provided they are at position 02.

Tier 1 Discussion Requires knowledge of the basis for emergency operating procedure steps. Meets Tier 1 requirements.

Technical Reference(s)

OI-15, EOP and EAL Clarifications Attached w/ Revision #

PPM 5.0.10, Flowchart Training Manual See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7784 - Given a list, identify the criteria that must be met to ensure that the existing rod pattern alone can always assure reactor shutdown.

Question Source: #: LO01786

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: 2009 Question Cognitive Level:

Justification for Cognitive Level

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-5 Requires memory of rod configuration to ensure reactor shutdown following a scram 10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

OI-15 Rev: Major: 032 Minor: N/A

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-5 Comments /

Reference:

PPM 5.0.10 Rev: Major: 023 Minor: N/A

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-6 Examination Outline Cross-reference: 6 Revision: 0 Date: 5/26/20 Tier: 1 Group: 1 K/A Number: 295016.2.4.31 Level of Difficulty: 2 RO Importance Rating: 4.2 K/A

Description:

Control Room Abandonment: Knowledge of annunciator alarms, indications, or response procedures.

CGS is in Mode 1.

A fire in the control room occurs.

The shift manager directs ABN-CR-EVAC, Control Room Evacuation and Remote Cooldown.

The control room is evacuated.

All immediate actions of ABN-CR-EVAC are complete.

RPV level is lowering.

What is the highest RPV level, as indicated at the Remote Shutdown Panel, where Emergency Depressurization is required?

A. -147 inches B. -150 inches C. -161 inches D. -186 inches Answer: A K/A Match:

Requires knowledge of indications that require actions during control room evacuation.

Explanation:

A. Correct. In accordance with the ABN-CR-EVAC flow chart, IF Div 1 125 Battery LE 108 VDC OR RPV Level LE -147 THEN Emergency Depressurization.

B. Incorrect. Plausible since this is the bottom of scale for MS-LI-10, RPV level indication on the Remote Shutdown Panel. However, ABN-CR-EVAC requires Emergency Depressurization LE -

147 inches.

C. Incorrect. Plausible since this is the highest RPV level that requires an ED during non-ATWS conditions in accordance with PPM 5.1.1, RPV Level Control, step L-11. However, ABN-CR-EVAC requires Emergency Depressurization LE - 147 inches.

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-6 D. Incorrect. Plausible since this is the highest RPV level that requires an ED during ATWS conditions in accordance with PPM 5.1.2, RPV Level Control - ATWS, step L-15. However, ABN-CR-EVAC requires Emergency Depressurization LE - 147 inches.

Tier 1 Discussion Requires knowledge of parameters that need to be met to perform a step in the abnormal procedure. Meets Tier 1 requirements.

Technical Reference(s)

ABN-CR-EVAC, Control Room Evacuation and Remote Cooldown Attached w/ Revision #

PPM 5.1.1, RPV Control See Comments / Reference PPM 5.1.2 - RPV Control - ATWS Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 16139 - Upon completion of this class, the student will be able to demonstrate the knowledge required to independently perform the Licensed Operator tasks associated with ABN-CR-EVAC.

Question Source: #: LO01585

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires memory of required level for ED in ABN-CR-EVAC 10 CFR Part 55 Content: 55.41 10

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-6 Comments /

Reference:

ABN-CR-EVAC Rev: Major: 042 Minor: N/A

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-6 Comments /

Reference:

PPM 5.1.1 Rev: Major: 22 Minor: N/A Comments /

Reference:

PPM 5.1.2 Rev: Major: 26 Minor: N/A

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-7 Examination Outline Cross-reference: 7 Revision: 0 Date: 10/22/20 Tier: 1 Group: 1 K/A Number: 295018.AK1.01 Level of Difficulty: 2 RO Importance Rating: 3.5 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on component/system operation.

CGS is operating in Mode 3.

RCC-P-1A & 1B are running.

Subsequently, RCC-P-1B trips on overcurrent.

The CRS enters ABN-RCC, Loss of RCC.

What actions should be taken?

A. Verify RCC-P-1C is running.

B. Re-open RCC-V-6, RCC Supply to RW/RB.

C. Stop the running Reactor Water Cleanup (RWCU) pump.

D. Periodically swap running CRD pumps to ensure they do not overheat.

Answer: A K/A Match:

Requires knowledge of system of system response to a partial loss of RCC.

Explanation:

A. Correct. When the breaker of any running RCC pump opens for any reason, either from stopping the pump or from the breaker tripping, the standby pump in Auto will start. RCC-V-6 will remain open since the standby RCC pump will start prior to the 10 second time delay for closing the valve on less than two running pumps. ABN-RCC, step 4.2.1 directs operators to verify that the standby RCC pump has started.

B. Incorrect. Plausible since ABN-RCC, step 4.2.2 directs operators to verify RCC-V-6 is open (RCC-V-6 will close with less than 2 RCC pump breakers closed), and only RCC-P-1A breaker is closed for a short period of time. Plausibility is enhanced if it is believed that RCC pump auto-start interlock is not active out of Mode 1. However, RCC-V-6 will remain open since the standby RCC pump will start prior to the 10 second time delay for closing the valve on less than two running pumps.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-7 C. Incorrect. Plausible since since ABN-RCC, step 4.2.3 directs operators to secure the running RWCU pump and isolate RWCU if RCC-V-6 is closed. Plausibility is enhanced if it is believed that RCC pump auto-start interlock is not active out of Mode 1. However, RCC-P-1C will start when the running pump breaker opens and RCC-V-6 will remain open.

D. Incorrect. Plausible since ABN-RCC, step 4.2.6 directs operators to swap CRD pumps if they are overheating due to a loss of RCC. However, RCC-P-1C will start when the running pump breaker opens. Additionally, RCC-V-6 will remain open since the standby pump will start in less than 10 seconds. Therefore, adequate CRD pump cooling will remain.

Tier 1 Discussion Requires knowledge of actions required by abnormal procedure for the conditions given in the question. Meets Tier 1 requirements.

Technical Reference(s)

SD000196, RCC System Description Attached w/ Revision #

ABN-RCC, Loss of RCC See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5706 - Explain the interlocks associated with the following components or system conditions, including setpoints:

a. RCC pump auto start

b. RCC Pump trips.

c. Makeup to Surge Tank RCC-V-48.

d. Radwaste/Rx Bldg Supply RCC-V-6 Question Source: #: LO00791

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires an understanding of RCC pump and valve interlocks and a knowledge of abnormal procedures used to mitigate a loss of an RCC pump.

10 CFR Part 55 Content: 55.41 7 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-7 Comments /

Reference:

SD000196 Rev: Major: 14 Minor: 2 Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-7 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-7 Comments /

Reference:

ABN-RCC Rev: Major: 006 Minor: 004 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 Examination Outline Cross-reference: 8 Revision: 1 Date: 10/22/20 Tier: 1 Group: 1 K/A Number: 295019.AK2.14 Level of Difficulty: 3 RO Importance Rating: 3.2 K/A

Description:

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following: Plant air systems CGS is in Mode 1.

A leak in the Containment Instrument Air (CIA) system occurs.

The crew is taking actions in accordance with ABN-CIA, Containment Instrument Air System Failure.

Current CIA Main Header Pressure (CIA-PI-20) is 136 psig and lowering at 1 psig every 5 minutes.

What actions should be taken?

In accordance with ABN-CIA, operators should (1) and (2) .

A. (1) Place one loop of RHR in Suppression Pool Cooling (2) lineup to supply CIA with Control and Service Air (CAS)

B. (1) Place one loop of RHR in Suppression Pool Cooling (2) start an immediate Reactor shutdown C. (1) Install ADS solenoid keys in H13-P628 or H13-P631 (2) lineup to supply CIA with Control and Service Air (CAS)

D. (1) Install ADS solenoid keys in H13-P628 or H13-P631 (2) start an immediate Reactor shutdown Answer: C K/A Match:

Requires knowledge of actions required to operate plant air systems on a loss of CIA.

Explanation:

A. Incorrect. Plausible since lining up CAS to supply CIA is an action required by ABN-CIA for the conditions given in the question stem. However, lining up RHR for Suppression Pool Cooling is only required after the reactor is manually scrammed due to an imminent closure of MSIVs Page 1 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 (immediate action step 4.1. Closure of MSIVs occur at a header pressure between 50-80 psig, and is not imminent for the conditions given in the question stem B. Incorrect. Plausible since both actions are required for specific conditions in ABN-CIA. However, lining up RHR for Suppression Pool Cooling is only required after the reactor is manually scrammed due to an imminent closure of MSIVs (immediate action step 4.1. Closure of MSIVs occur at a header pressure between 50-80 psig, and is not imminent for the conditions given in the question stem. Additionally, a reactor shutdown is only required if CIA Main Header pressure drops to 80 psig (step 4.4).

C. Correct. In accordance with ABN-CIA, step 4.3, if Main Header pressure (CIA-PI-20) approaches 135 psig, operators should install ADS solenoid keys and supply CIA with CAS.

D. Incorrect. Plausible since installing ADS Solenoid Keys is a required action for the conditions given in the question stem. However, a reactor shutdown is only required if CIA Main Header pressure drops to 80 psig (step 4.4).

Tier 1 Discussion Requires knowledge of ABN actions. Meets Tier 1 requirements.

Technical Reference(s)

ABN-CIA, Containment Instrument Air System Failure Attached w/ Revision #

See Comments / Reference Proposed references to be provided during examination: N/A Learning Objective: 5146 - State the four (4) sources for the CIA and when each source would be used.

Question Source: #: Bank #

LO01870 (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Candidate must have knowledge of the actions required by ABN-CIA for specific plant conditions.

10 CFR Part 55 Content: 55.41 7 Page 2 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 Comments /

Reference:

ABN-CIA Rev: Major: 007 Minor: 001 Page 3 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 Page 4 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 Comments /

Reference:

Question LO01870 Rev: Major: Maj Minor: Min Page 5 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-8 Page 6 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-9 Examination Outline Cross-reference: 9 Revision: 0 Date: 9/8/20 Tier: 1 Group: 1 K/A Number: 295021.AA2.07 Level of Difficulty: 4 RO Importance Rating: 2.9 K/A

Description:

Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING:

Reactor recirculation flow CGS is in Mode 4.

The reactor has been shutdown for 3 days.

RHR Loop B is in shutdown cooling.

RHR-P-2A is tagged out for emergent repairs.

Both RRC pumps are off, but available.

RHR-P-2B trips due to a failure of the pump breaker.

The CRS enters ABN-RHR-SDC-LOSS, Loss of Shutdown Cooling.

Maintenance reports that it will take 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months to replace the RHR-P-2B breaker.

What is the preferred method for restoring core circulation until RHR-P-2B is restored?

A. Start a RRC pump and monitor RPV temperature.

B. Raise RPV level to GE 60 inches to enhance natural circulation.

C. Makeup to the RPV with the CRD system and letdown to the hotwell through the RWCU system.

D. Bleed steam through the main steam drain lines while maintaining RPV level with Condensate Booster Pumps.

Answer: A K/A Match:

Requires evaluation of Explanation:

A. Correct. With the plant in Mode 4, Technical Specification LCO 3.4.10 requires forced core circulation. A running RRC pump is one method to provide forced circulation. ABN-RHR-SDC-LOSS specifies this as the preferred method when both RHR pumps are unavailable.

B. Incorrect. Plausible since step 4.3 of ABN-RHR-SDC-LOSS directs raising RPV level to GE 60 inches to enhance natural circulation for if no RHR pumps or RRC pumps can be restored.

However, this is only performed if both RHR pumps or a single RRC pump cannot be started.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-9 C. Incorrect. Plausible since this is a method for additional decay heat removal listed in ABN-RHR-SDC-LOSS. However, this is not the preferred method to restore forced circulation.

D. Incorrect. Plausible since this is a method for alternate decay heat removal listed in ABN-ADHR.

However, this is not the preferred method to restore forced circulation.

Technical Reference(s)

ABN-RHR-SDC-LOSS, Loss of Shutdown Cooling Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7728 - Describe the physical connection and/or cause-and-effect relationships between the RHR system and the following: e. RRC Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing the information given in the question stem with an understanding of the methods available for forced circulation along with the preferred method of ABN-RHR-SDC-LOSS 10 CFR Part 55 Content: 55.41 10 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-9 Comments /

Reference:

ABN-RHR-SDC-LOSS Rev: Major: 7 Minor: N/A Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-9 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-10 Examination Outline Cross-reference: 10 Revision: 0 Date: 10/28/20 Tier: 1 Group: 1 K/A Number: 295023.AK3.01 Level of Difficulty: 2 RO Importance Rating: 3.6 K/A

Description:

Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS :

Refueling floor evacuation CGS is in Mode 5.

A core shuffle is in progress.

As a fuel bundle is being removed from the core, it impacts a structural component.

Gas bubbles are visible from the fuel bundle.

What action should be taken?

A. Evacuate the refueling floor to minimize potential dose to personnel.

B. Move the fuel bundle to the spent fuel pool to minimize contamination of the RPV.

C. Move the fuel bundle to its original location to minimize potential for further damage.

D. Perform a subcritical check to ensure inadvertent criticality is not occurring.

Answer: A K/A Match:

Requires knowledge of the reasons for actions in ABN-FUEL-HAND, Damage While Handling Fuel.

Explanation:

A. Correct. One of the entry conditions for ABN-FUEL-HAND, Damage While Handling Fuel, is a fuel bundle impact with another object and visible gas bubbles. Action 4.1 is NOTIFY the Refuel Floor Supervisor/Spent Fuel Pool Supervisor to evacuate the refuel. In accordance with the ABN Bases, this step is performed to minimize the dose to personnel on the refueling floor.

B. Incorrect. Plausible since PPM 6.3.2, Fuel shuffling and/or Offloading and Reloading, Precaution and Limitation 5.19 allows a fuel bundle to be placed in a temporary location if a difficulty is encountered during fuel moves. However, for the conditions given, ABN-FUEL-HAND should be entered and the refueling floor should be evacuated.

C. Incorrect. Plausible since PPM 6.3.2, Fuel shuffling and/or Offloading and Reloading, Precaution and Limitation 5.21 allows a fuel bundle to be placed in its original location if the correct fuel move cannot be made. However, for the conditions given, ABN-FUEL-HAND should be entered and the refueling floor should be evacuated.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-10 D. Incorrect. Plausible since a subcritical check is performed if an unexpected SRM count rate is encountered during a fuel shuffle (see PPM 6.3.2, section 2.2) However, for the conditions given, ABN-FUEL-HAND should be entered and the refueling floor should be evacuated.

Tier 1 Discussion Requires knowledge of AOP entry conditions and actions required. Meets Tier 1 requirements.

Technical Reference(s)

ABN-FUEL-HAND, Damage while Handling Fuel Attached w/ Revision #

PPM-6.3.2, Fuel Shuffling and/or Offloading and Reloading See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8839 - Describe the operator action when fuel damage while refueling is indicated.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must synthesize the conditions given in the question stem with a knowledge of ABN entry conditions along with required ABN actions and reasons for these actions.

10 CFR Part 55 Content: 55.41 13 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-10 Comments /

Reference:

ABN-FUEL-HAND Rev: Major: 004 Minor: 002 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-10 Comments /

Reference:

PPM 6.3.2 Rev: Major: 025 Minor: N/A Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-11 Examination Outline Cross-reference: 11 Revision: 0 Date: 6/10/20 Tier: 1 Group: 1 K/A Number: 295024.EK3.04 Level of Difficulty: 2 RO Importance Rating: 3.7 K/A

Description:

Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE :

Emergency depressurization CGS is in Mode 1.

An event occurs causing Drywell pressure and Wetwell pressure to rise.

The CRS enters PPM 5.2.1, Primary Containment Control.

Based on Drywell and Wetwell pressure, when is an Emergency Depressurization (ED) required?

An ED is required when Wetwell pressure cannot be restored and maintained below A. Drywell Spray Initiation Limit (DSIL).

B. Pressure Suppression Pressure (PSP).

C. Decay Heat Removal Pressure (DHRP).

D. Primary Containment Pressure Limit (PCPL).

Answer: B K/A Match:

Requires knowledge of the reason for performing an emergency depressurization with high primary containment pressure.

Explanation:

A. Incorrect. Plausible since DSIL is used in the Primary Containment pressure leg to determine if drywell spray may be initiated. However, it is not used to determine if an ED is required.

B. Correct. In accordance with PPM 5.2.1, step P-12, when Wetwell pressure cannot be restored and maintained below PSP an ED is required.

C. Incorrect. Plausible since DHRP is determined by the difference in pressure between the RPV and primary containment and as Drywell and Wetwell pressure rise, the plant is closer to the DHRP limit. However, DHRP is not used to determine when an ED is required.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-11 D. Incorrect. Plausible since exceeding PCPL requires actions in the primary containment pressure leg of PPM 5.2.1. (step P-13). However, PCPL is not used to determine when an ED is required.

Exceeding PCPL requires venting primary containment.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control Attached w/ Revision #

PPM 5.0.10, Flowchart Training Manual See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: Learning Objective.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the conditions that require an emergency depressurization with high primary containment pressure.

10 CFR Part 55 Content: 55.41 5 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-11 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-11 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-11 Comments /

Reference:

PPM 5.0.10 Rev: Major: 28 Minor: N/A Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-12 Examination Outline Cross-reference: 12 Revision: 0 Date: 10/28/20 Tier: 1 Group: 1 K/A Number: 295025.EA2.06 Level of Difficulty: 3 RO Importance Rating: 3.7 K/A

Description:

Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Reactor water level CGS is in Mode 1.

The reactor automatically scrams on high RPV pressure.

Current plant conditions:

Reactor pressure is 1065 psig, up slow.

RPV level is +40 inches.

The crew is taking actions in accordance with PPM 5.1.1, RPV Control.

The CRS has directed CRO1 to Restore and Maintain RPV water level +13 inches to +54 inches in accordance with PPM 5.1.1, step L-3.

For current plant conditions, which of the following RPV level instruments will indicate the most accurate RPV level.

A. Shutdown (MS-LI-605)

B. Upset (RFW-LR-608)

C. Compensated Fuel Zone (MS-LR-615)

D. Uncompensated Fuel Zone (MS-LI-610)

Answer: B K/A Match:

Requires knowledge of reactor water level accuracy with high reactor pressure.

Explanation:

A. Incorrect. Plausible if it is believed that the Shutdown range is calibrated for higher reactor pressure. However, it is calibrated for 0 psig while other ranges are calibrated for 1000 psig.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-12 B. Correct. The Upset Range is calibrated for 1000 psig, while the other distractors are calibrated for 0 psig.

C. Incorrect. Plausible since this indication is density compensated and is valid for all pressure ranges. However, it is calibrated for 0 psig.

D. Incorrect. Plausible if it is believed that the Shutdown range is calibrated for higher reactor pressure. However, it is calibrated for 0 psig while other ranges are calibrated for 1000 psig.

Tier 1 Discussion Requires knowledge of which RPV level indicator to use while performing EOP actions. Meets Tier 1 requirements.

Technical Reference(s)

PPM 5.1.1, RPV Control Attached w/ Revision #

SD000126, Nuclear Boiler Instrumentation System Description See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5582 - List the calibration conditions and nominal ranges for each of the five ranges of level instruments.

a. NARROW RANGE

c. WIDE RANGE

d. UPSET RANGE

e. SHUTDOWN RANGE

f. FUEL ZONE RANGE Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of RPV level instrumentation calibration conditions.

10 CFR Part 55 Content: 55.41 7 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-12 Comments /

Reference:

PPM 5.1.1 Rev: 22 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-12 Comments /

Reference:

SD000126 Rev: Major: 13 Minor: 2 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-13 Examination Outline Cross-reference: 13 Revision: 0 Date: 6/11/20 Tier: 1 Group: 1 K/A Number: 295026.EA2.03 Level of Difficulty: 2 RO Importance Rating: 3.9 K/A

Description:

Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor pressure CGS is in Mode 1.

The CRS has entered PPM 5.2.1 due to high Wetwell temperature.

Current conditions:

Wetwell temperature: 210°F.

Wetwell level: 22 feet.

What is the lowest RPV pressure which would result in exceeding HCTL?

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-13 A. 400 psig B. 500 psig C. 650 psig D. 700 psig Answer: B K/A Match:

Requires ability to interpret reactor pressure with respect to HCTL with high suppression pool temperature.

Explanation:

A. Incorrect. Plausible if it is believed that the unsafe region is below the line. However, the unsafe region is above the line.

B. Correct. Since the given Wetwell level falls between the lines given on the HCTL graph, the most conservative line (i.e. lowest Wetwell level) should be used. With Wetwell temperature at 210°F, the 19.2 foot line intersects at approximately 490 psig. Therefore, 500 psig is the lowest RPV pressure given that exceeds HCTL.

C. Incorrect. Plausible since 650 psig is the value of RPV pressure if attempting to interpolate the value of RPV pressure for a Wetwell level of 22 feet. However, PPM 5.0.10 states When using either HCTL graphs, if WW level is not on one of the lines on the graph, use the next line below (move in the direction to make the unsafe area larger) to determine HCTL.

D. Incorrect. Plausible if the 23 foot line is used. However, the most conservative line (19.2 feet) should be used.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control Attached w/ Revision #

PPM 5.0.10, Flowchart Training Manual See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8302 - Given plant conditions and the Heat Capacity Temperature Limit Curve, determine the current operating point on the Curve within 2.5 degrees and 25 psig.

Question Source: #: LO02944 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-13

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires a knowledge of the usage rules for the HCTL graph and the ability to read the HCTL graph.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-13 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-13 Comments /

Reference:

PPM 5.0.10 Rev: Major: 23 Minor: N/A Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-14 Examination Outline Cross-reference: 14 Revision: 0 Date: 8/7/20 Tier: 1 Group: 1 K/A Number: 295028.2.4.21 Level of Difficulty: 2 RO Importance Rating: 4.0 K/A

Description:

High drywell temperature: 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.

CGS is in Mode 1.

An event causes a high drywell temperature condition.

The CRS enters PPM 5.2.1, Primary Containment Control.

In accordance with PPM 5.2.1, when is the crew required to perform an Emergency Depressurization (ED) and what is the reason for this requirement?

An ED is required to be performed when drywell temperature cannot be restored and maintained below (1) . This is performed since containment parameters cannot be maintained below (2) .

A. (1) 285°F (2) Heat Capacity Temperature Limit (HCTL)

B. (1) 285°F (2) Automatic Depressurization System (ADS) design temperature C. (1) 330°F (2) Heat Capacity Temperature Limit (HCTL)

D. (1) 330°F (2) Automatic Depressurization System (ADS) design temperature Answer: D K/A Match:

Requires knowledge of parameter that requires action to maintain containment system safety function.

Explanation:

A. Incorrect. Plausible since 285°F is the best practice temperature in PPM 5.2.1 to proceed with the Drywell High Temperature leg. Plausibility is enhanced since an ED is performed on Wetwell Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-14 temperature when HCTL is exceeded. However, these parameters are not applicable to ED on Drywell temperature.

B. Incorrect. Plausible since the reason of performing an ED due to Drywell Temperature is correct.

However, the ED is performed at 330°F.

C. Incorrect. Plausible since an ED on high Drywell Temperature is performed if drywell temperature cannot be reduced and maintained below 330°F. However, the reason listed is for an ED on Wetwell Temperature vice Drywell Temperature.

D. Correct. In accordance with PPM 5.2.1, step DT-8, an ED is required when drywell temperature cannot be restored and maintained below 330°F. PPM 5.0.10, Flowchart Training manual states that this ED is performed when drywell temperature cannot be restored and maintained below the ADS design temperature.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control Attached w/ Revision #

PPM 5.0.10, Flowchart Training Manual See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: Learning Objective.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the requirements for performing an ED on high drywell temperature and the reason for this requirement.

10 CFR Part 55 Content: 55.41 10 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-14 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-14 Comments /

Reference:

PPM 5.0.10 Rev: Major: 022 Minor: 000 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-15 Examination Outline Cross-reference: 15 Revision: 0 Date: 11/23/20 Tier: 1 Group: 1 K/A Number: 295030.EK1.03 Level of Difficulty: 3 RO Importance Rating: 3.8 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Heat capacity CGS is in Mode 1.

An earthquake occurs causing wetwell level to lower.

The CRS enters PPM 5.2.1, Primary Containment Control.

Drywell pressure is 0.6 psig and steady.

Wetwell pressure is 0.1 psig and steady.

Wetwell temperature is 70°F and steady.

Wetwell level is 19 feet, 1 inch, down fast.

(1) What action is required?

(2) What is the basis for this action?

The crew must perform an (1) . This action is performed (2) .

A. (1) anticipated RPV depressurization per PPM 5.1.1, RPV Control.

(2) since the RPV is not allowed to be at pressure when pressure suppression capability is unavailable B. (1) anticipated RPV depressurization per PPM 5.1.1, RPV Control.

(2) to ensure equipment in the wetwell necessary for safe shutdown will remain operational C. (1) emergency depressurization per PPM 5.1.3, Emergency RPV Depressurization (3) since the RPV is not allowed to be at pressure when pressure suppression capability is unavailable D. (1) emergency depressurization per PPM 5.1.3, Emergency RPV Depressurization (2) to ensure equipment in the wetwell necessary for safe shutdown will remain operational Answer: C K/A Match:

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-15 Requires knowledge of the operational requirements on a loss or potential loss of heat capacity due to low suppression pool level.

Explanation:

A. Incorrect. Plausible since the required action is due to the potential loss of the pressure suppression capability for an emergency depressurization (ED). However, an ED is required since wetwell level is below 19 feet, 2 inches.

B. Incorrect. Plausible since an ED is required to ensure wetwell equipment remains operational due to wetwell temperature. However, for low wetwell level, an ED is required due to the imminent loss of the pressure suppression function.

C. Correct. In accordance with PPM 5.2.1, Primary Containment Control, an ED is required when wetwell level cannot be maintained above 19 feet, 2 inches. PPM 5.0.10, Flowchart Training Manual states that the reason for ED is that the RPV may not remain at pressure when pressure suppression capability is unavailable.

D. Incorrect. Plausible since an ED is required. However, an ED is required due to the imminent loss of the pressure suppression function.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control Attached w/ Revision #

PPM 5.0.10, Flowchart Training Manual See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11150 - Given plant conditions and EOP flowcharts, evaluate plant conditions and determine the appropriate actions according to EOP 5.2.1.

Question Source: #: LO03417

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-15 Justification for Cognitive Level Requires synthesizing information in question stem with a knowledge of the actions required along with the reason for performing these actions.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

PPM 5.1.2 Rev: Major: 28 Minor: N/A Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-15 Comments /

Reference:

PPM 5.0.10 Rev: Major: 22 Minor: 000 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 Examination Outline Cross-reference: 16 Revision: 1 Date: 11/23/20 Tier: 1 Group: 1 K/A Number: 295031.EK1.02 Level of Difficulty: 4 RO Importance Rating: 3.8 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL: Natural circulation CGS is in Mode 1.

A LOCA occurs.

A manual reactor scram is inserted.

Control rods fail to insert.

The crew is taking actions in accordance with PPM 5.1.2, RPV Control - ATWS.

The crew is controlling level -80 inches to -140 inches.

At step L-7, the CRS determines that Level/Power conditions exist and directs lowering RPV level.

Why is RPV level lowered at step L-8 of the procedure?

RPV level is lowered to reduce core Page 1 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 A. void fraction.

B. boron mixing.

C. natural circulation flow.

D. level oscillations.

Answer: C K/A Match:

Requires understanding of the effect on natural circulation flow when RPV level is lowered.

Explanation:

A. Incorrect. Plausible since void fraction is changed when RPV level is lowered. However, void fraction is increased to lower reactor power.

B. Incorrect. Plausible since actions to promote boron mixing should be taken to ensure that the entire core will shut down. However, lowering RPV level during Level/Power conditions is performed to reduce natural circulation flow, increase void fraction, and reduce reactor power.

Boron stratification is increased when natural circulation flow is lowered. Additionally, boron mixing should be promoted to ensure that the entire core feels the effects of boron addition.

C. Correct. If Level/Power conditions exist, heat could be rejected to the wetwell at a rate in excess of that which can be removed by the wetwell cooling system. This could result in containment over pressurization and loss of containment. Lowering RPV level reduces natural circulation flow.

This reduces the amount of steam that is removed from the RPV, which increases void fraction and adds negative reactivity to the core. This will cause reactor power to lower and reduce the amount of energy that is rejected to containment.

D. Incorrect. Plausible since the initial reduction of RPV level in step L-6 is performed to reduce core inlet subcooling and reduce neutron flux oscillations caused by thermal-hydraulic instabilities.

However, at step L-7, level reduction is performed to reduce natural circulation flow. Additionally, PPM 5.0.10 states that reducing RPV level significantly below normal level may cause level and power oscillations.

Tier 1 discussion Question requires knowledge of reason for performing EOP step. Meets Tier 1 requirements.

Technical Reference(s)

PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

PPM 5.1.2, RPV Control - ATWS See Comments / Reference Tech Ref 3 Tech Ref 4 Page 2 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 Proposed references to be provided during examination: N/A Learning Objective: 8120 - Given a list, identify the statements that describe the purpose of maintaining RPV water level between the Minimum Steam Cooling RPV Water Level and either +54 inches or LL following emergency depressurization during an ATWS.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information given in the question stem with a knowledge of 10 CFR Part 55 Content: 55.41 1 Comments /

Reference:

PPM 5.1.2 Rev: Major: 26 Minor: N/A Page 3 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 Comments /

Reference:

PPM 5.0.10 Rev: Major: 023 Minor: N/A Page 4 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 Page 5 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 Page 6 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 Page 7 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-16 Page 8 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-17 Examination Outline Cross-reference: 17 Revision: 0 Date: 6/22/20 Tier: 1 Group: 1 K/A Number: 295037.EK3.06 Level of Difficulty: 3 RO Importance Rating: 3.8 K/A

Description:

Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : Maintaining heat sinks external to the containment CGS is in Mode 1.

An event occurs that requires a reactor scram.

Control rods fail to insert.

The crew is taking actions in accordance with PPM 5.1.2, RPV Control - ATWS.

Plant conditions:

SLC has failed to inject.

Reactor power is 15% and stable.

Outboard MSIVs automatically closed due to a loss of CAS pressure.

CAS pressure to MSIVs has been restored.

The CRS directs re-opening the MSIVs.

With the current plant conditions, what limit will first be exceeded if MSIVs are not opened?

MSIVs are opened to discharge heat energy to the main condenser to prevent exceeding...

A. SRV Tail Pipe Level Limit (SRVTPLL).

B. Pressure Suppression Pressure (PSP).

C. Heat Capacity Temperature Limit (HCTL).

D. Primary Containment Pressure Limit (PCPL).

Answer: C K/A Match:

Requires knowledge of reason for restoring an external heat sink during an ATWS Explanation:

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-17 A. Incorrect. Plausible if it is believed that wetwell level will be the most limiting parameter for the conditions given in the stem. However, for the conditions given, an external heat sink is required to maintain the wetwell below HCTL.

B. Incorrect. Plausible since adding heat to primary containment may cause containment pressure to rise. However, for the conditions given, an external heat sink is required to maintain the wetwell below HCTL.

C. Correct. In accordance with PPM 5.1.2, RPV Control - ATWS, override P-4, MSIVs should be opened when boron injection is required. PPM 5.0.10 gives the reason for this: To stabilize and control RPV pressure, the reactor steam generation rate must remain within the capacity of systems designed to remove the steam from the RPV. With the reactor not shutdown, the amount of steam that may have to be released could be substantial. If this heat energy is discharged to the Wetwell, the HCTL could be reached in a very short time. Therefore, utilization of the main condenser as a heat sink for this energy is of sufficient importance to warrant opening the MSIVs D. Incorrect. Plausible since maintaining primary containment is the goal of opening the MSIVs.

However, HCTL will be exceeded first.

Technical Reference(s)

PPM 5.1.2, RPV Control - ATWS Attached w/ Revision #

PPM 5.0.10, Flowchart Training Manual See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8169 - Describe why re-opening MSIVs when boron injection is required, is beneficial during an ATWS.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of reason for an action in the EOPs.

10 CFR Part 55 Content: 55.41 10 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-17 Comments /

Reference:

PPM 5.1.2 Rev: Major: 26 Minor: N/A Comments /

Reference:

PPM 5.0.10 Rev: Major: 023 Minor: N/A Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-17 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-18 Examination Outline Cross-reference: 18 Revision: 1 Date: 10/21/20 Tier: 1 Group: 1 K/A Number: 295038.EA1.01 Level of Difficulty: 3 SRO Importance Rating: 3.9 K/A

Description:

Ability to operate and/or monitor the following as they apply to HIGH OFF-SITE RELEASE RATE: Stack-gas monitoring system.

CGS is in Mode 1.

An event causes the following readings on the Reactor Building Exhaust Plenum Radiation Monitors:

REA-RIS-609A: 11 mr/hr, steady

REA-RIS-609B: 11 mr/hr, steady

REA-RIS-609C: 14 mr/hr, steady

REA-RIS 609D: Downscale The CRS enters PPM 5.3.1, Secondary Containment Control.

With respect to Reactor Building Ventilation, what should the operator verify?

In accordance with PPM 5.3.1, verify ROA-V-1, Reactor Building Supply Inboard Isolation, is (1) , and ROA-V-2, Reactor Building Supply Outboard Isolation, is (2) .

A. (1) closed (2) closed B. (1) closed (2) open C. (1) open (2) closed D. (1) open (2) open Answer: B K/A Match:

Requires knowledge of the expected outcome for stack gas monitoring system conditions when verifying equipment in accordance with PPM 5.3.1, Secondary Containment Control.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-18 Explanation:

A. Incorrect. Plausible since ROA-V-2 will be closed. However, ROA-V-1 will remain open B. Correct. REA-RIS-609C & D will provide a close signal to the RB HVAC inboard isolation valves when both monitors are > Hi-Hi level (13 mr/hr), Downscale, or a combination of either condition.

C. Incorrect. Plausible if it is believed that REA-RIS-609C & D operate the outboard valves.

However, these monitors provide input to the inboard valves.

D. Incorrect. Plausible since ROA-V-2 will be open. Plausibility is enhanced if it is believed that a monitor downscale will not provide a close signal. However, for the conditions given, ROA-V-1 will close.

Tier 1 Discussion: This question requires knowledge of the expected system response while performing step SC-1 of EOP PPM 5.2.1, Secondary Containment Control.

Technical Reference(s)

SD000147, Process Radiation Monitoring Attached w/ Revision #

PPM 5.3.1, Secondary Containment Control See Comments / Reference Proposed references to be provided during examination: N/A Learning Objective: 5647 State the automatic actions associated with each of the following gaseous and liquid stream Process Radiation Monitors upon sensing high radiation levels: g. Reactor Building Exhaust Plenum RMS Question Source: #:

(Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must demonstrate a knowledge of the conditions required to isolate RB HVAC along with a knowledge of the isolation signals provided by individual radiation monitors.

10 CFR Part 55 Content: 55.41 11 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-18 Comments /

Reference:

SD000147 Rev: Major: 15 Minor: 001 Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor: N/A Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-18 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Examination Outline Cross-reference: 19 Revision: 2 Date: 11/23/20 Tier: 1 Group: 1 K/A Number: 600000.2.2.44 Level of Difficulty: 3 SRO Importance Rating: 4.2 K/A

Description:

Fire Protection: Ability to interpret control room indications to verify the status and operation of a system and understand how operator actions and directives affect plant and system conditions.

CGS is in Mode 1.

The following annunciator alarms in the control room:

4.FCP1.2-2 4.FCP2.4-2 Field operators report a fire in switchgear SH-5.

The crew enters ABN-Fire.

Operators deenergize SH-5 in accordance with Pre-Fire Plan PFP-TG-471.

One minute later, operators verify that at least one fire pump is running in accordance with ABN-Fire, step 4.7.

Fire Main Pressure is 119 psig, down slow.

Which of the following fire pump configurations is correct for the conditions given?

A.

B.

Page 1 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 C.

D.

Answer: B K/A Match:

Requires interpretation of plant indications and determining how the Fire Protection system will respond when SH-5 is isolated.

Explanation:

A. Incorrect. Plausible since the power supply to FP-P-2A is MC-5N, which is powered from SH-5.

Although FP-P-2A should automatically start when Fire Main Pressure is LE 120 psig, it will not start since it is deenergized. However, control power for FP-P-1 (diesel driven) is MC-5N. With a loss of SH-5, MC-5N is lost. FP-P-1 will automatically start 20 seconds after a loss of control power.

B. Correct. The power supply to FP-P-2A is MC-5N, which is powered from SH-5. Although FP-P-2A should automatically start when Fire Main Pressure is LE 120 psig, it will not start since it is deenergized. However, control power for FP-P-1 (diesel driven) is MC-5N. With a loss of SH-5, MC-5N is lost. FP-P-1 will automatically start 20 seconds after a loss of control power. FP-P-2B automatically starts after fire main pressure is LE 110 psig C. Incorrect. Plausible if it is believed that FP-P-2A does not lose power with a loss of SH-5.

Plausibility is enhanced since fire main pressure is below the auto start setpoint for FP-P-2A.

However, FP-P-2A does lose power and cannot start. Additionally, control power for FP-P-1 (diesel driven) is MC-5N. With a loss of SH-5, MC-5N is lost. FP-P-1 will automatically start 20 seconds after a loss of control power.

D. Incorrect. Plausible since control power for FP-P-1 (diesel driven) is MC-5N. With a loss of SH-5, MC-5N is lost. FP-P-1 will automatically start 20 seconds after a loss of control power. Plausibility is enhanced if it is believed that FP-P-2B auto setpoint is at 120 psig. However, the auto start setpoint for FP-P-2B is 110 psig.

Tier 1 Discussion The question requires candidates to know the actions necessary to isolate equipment in accordance with the Pre-Fire Plan and how to validate these actions using pump indications.

Page 2 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Technical Reference(s)

SD000177, Fire Protection System Description Attached w/ Revision #

SOP-ELEC-SH5-MAINT, Removing SH-5 From Service See Comments / Reference ABN-ELEC-SH5, SH-5 Distribution System Failures PFP-TG-471, Turbine Generator 471 Pre-Fire Plan Proposed references to be provided during examination: N/A Learning Objective: 12271 - Explain the function and operation of the following Fire Protection System components, including any automatic features or interlocks:

a. Diesel Fire Pumps, FP-P-1, FP-P-110

b. Electric Fire Pumps, FP-P-2A, FP-P-2B Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires interpreting plant conditions given in the question stem to determine condition of the electrical distribution system and a knowledge of how plant conditions affect the Fire Protection system.

10 CFR Part 55 Content: 55.41 4 Page 3 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Comments /

Reference:

SD000177 Rev: Major: 16 Minor: 4 Page 4 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Page 5 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Comments /

Reference:

SOP-ELEC-SH5-MAINT Rev: Major: 012 Minor: 003 Page 6 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Comments /

Reference:

ABN-ELEC-SH5 Rev: Major: 000 Minor: 004 Page 7 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-19 Comments /

Reference:

PFP-TG-471 Rev: Major: 004 Minor: 001 Page 8 of 8

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-20 Examination Outline Cross-reference: 20 Revision: 0 Date: 7/22/20 Tier: 1 Group: 1 K/A Number: 700000.AA1.05 Level of Difficulty: 3 SRO Importance Rating: 3.9 K/A

Description:

Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Engineered safety features CGS is in Mode 1.

DG-2 is paralleled to the grid to complete OSP-ELEC-S702, DG2 Semi-Annual Operability Test.

The BPA dispatcher informs the control room that the 500 kV system condition is degraded.

The CRS enters ABN-ELEC-GRID, Degraded Off Site Power Grid.

How should the crew operate the emergency diesel generators?

A. Startup DG-1 and parallel it to the grid.

B. Shutdown DG-2 and place it in standby.

C. Unload DG-2 and start DG-1. Run both DGs unloaded.

D. Power SM-7 and SM-8 from their respective DGs and divorce both buses from off site power.

Answer: B K/A Match:

Requires knowledge of how to operate ESF equipment (diesel generators) with a grid disturbance.

Explanation:

A. Incorrect. Plausible since both DGs are picking up ESF bus loads when paralleled to the grid.

However, it is not desirable to parallel the DGs with an unstable grid.

B. Correct. with a DG synchronized to an unstable grid, the possibility of losing the DG as a reliable power source increases. ABN-ELEC-GRID, Degraded Off Site Power Grid, step 4.1.13, directs If any diesel generator is synchronized to the grid, then separate the diesel from the grid, and return the diesel to standby per the applicable procedure.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-20 C. Incorrect. Plausible since DGs running unloaded are capable of picking up ESF bus loads.

However, the DGs should not be run unloaded for long periods, and ABN-ELEC-GRID directs the DGs to be placed in standby D. Incorrect. Plausible since the ESF buses would be powered from reliable power supplies.

However, ABN-ELEC-GRID directs placing both DGs in standby.

Technical Reference(s)

ABN-ELEC-GRID, Degraded Off Site Power Grid Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 15748 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-ELEC-GRID.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing information given in the question stem with a knowledge of action required per ABN-ELEC-GRID along with an understanding that the DGs are not to be run unloaded for a prolonged period of time.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-20 Comments /

Reference:

ABN-ELEC-GRID Rev: Major: 011 Minor: N/A Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 Examination Outline Cross-reference: 21 Revision: 0 Date: 11/23/20 Tier: 1 Group: 2 K/A Number: 295009.AK1.02 Level of Difficulty: 2 RO Importance Rating: 3.0 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to LOW REACTOR WATER LEVEL: Recirculation pump net positive suction head CGS is in Mode 1.

A loss of Reactor Feed Water Pump, RFW-P-1A occurs.

RPV Level is +25 inches, down slow.

The following annunciator alarms:

4.602.A6.3-3: LOOP A DELTA T CAVITATION LIMIT Operators have determined that the low T condition is valid.

How should operators respond to this event?

IF the low T condition remains for GTE (1) , operators should verify RRC pumps automatically (2) .

A. (1) 3 minutes (2) runback to 15 Hz B. (1) 3 minutes (2) trip C. (1) 10 minutes (2) runback to 15 Hz D. (1) 10 minutes (2) trip Answer: C Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 K/A Match:

Requires knowledge of the operational limits imposed to prevent a loss of NPSH to RRC pumps.

Explanation:

A. Incorrect. Plausible since ARP 4.602.A6.3-3, requires operators to manually runback RRC pumps with a valid low T condition. However, RRC pumps will not automatically runback until the valid signal is present for GTE 10 minutes B. Incorrect. Plausible since actions should be taken if the low delta-T condition remains for GT 3 minutes. However, RRC pumps are not tripped, they are runback to 15 Hz.

C. Correct. In accordance with ARP 4.602.A6.3-3, if a valid low T signal is present for GTE 10 minutes, operators should verify that RRC pumps automatically runback to 15 Hz.

D. Incorrect. Plausible since a runback will automatically occur at GT 10 minutes. However, RRC pumps will not automatically trip.

Tier 1 Discussion Requires knowledge of actions required by Alarm Response Procedures. Meets Tier 1 requirements.

Technical Reference(s)

Final Safety Analysis Report (FSAR) 4.602.A6.3-3, Alarm Response for LOOP A DELTA T CAVITATION Attached w/ Revision #

LIMIT See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 9686 - Given an initial operating condition, describe the response of the RRFC system to DT LT 10.7°F between RPV Steam Dome and RRC Suction.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must have knowledge of the actions required for a low T condition and the time allowed to perform this action.

Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 10 CFR Part 55 Content: 55.41 2 Comments /

Reference:

FSAR Rev: Amendment 63 Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 Comments /

Reference:

4.602.A6.3-3 Rev: Major: 033 Minor: 002 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-21 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-22 Examination Outline Cross-reference: 22 Revision: 0 Date: 7/24/20 Tier: 1 Group: 2 K/A Number: 295010.AA2.02 Level of Difficulty: 2 SRO Importance Rating: 3.8 K/A

Description:

Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: Drywell Pressure.

CGS is in Mode 1.

A LOCA occurs.

The crew is taking actions in accordance with PPM 5.1.1, RPV Control, and PPM 5.2.1, Primary Containment Control.

Current plant conditions:

RHR-B is spraying the wetwell.

RHR-A is spraying the drywell.

Drywell pressure: 16 psig, down slow.

Wetwell pressure 13 psig, down slow.

When should drywell spray be secured?

Secure drywell spray when drywell pressure A. goes below 0 psig.

B. reaches 1.68 psig.

C. goes below 12 psig.

D. is not within the Drywell Spray Initiation Limit (DSIL).

Answer: B K/A Match:

Requires ability to interpret drywell pressure in relation to securing containment sprays.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-22 Explanation:

A. Incorrect. Plausible since containment sprays are required to be secured prior to going below 0 psig. However, in accordance with PPM 5.2.1, Primary Containment Control, PPM 5.0.10, Flowchart Training Manual, and OI-15, EO and EAL Clarifications, drywell spray is secured when drywell pressure reaches 1.68 psig.

B. Correct. In accordance with PPM 5.2.1, Primary Containment Control, PPM 5.0.10, Flowchart Training Manual, and OI-15, EO and EAL Clarifications, drywell spray is secured when drywell pressure reaches 1.68 psig.

C. Incorrect. Plausible since drywell spray is not initiated until drywell pressure exceeds 12 psig in accordance with PPM 5.2.1, step P-7. This suggests that drywell spray is not needed with drywell pressure less than 12 psig. However, in accordance with PPM 5.2.1, Primary Containment Control, PPM 5.0.10, Flowchart Training Manual, and OI-15, EO and EAL Clarifications, drywell spray is secured when drywell pressure reaches 1.68 psig.

D. Incorrect. Plausible since drywell spray is not initiated if drywell parameters are not within DSIL in accordance with PPM 5.2.1, step P-8. However, in accordance with PPM 5.2.1, Primary Containment Control, PPM 5.0.10, Flowchart Training Manual, and OI-15, EO and EAL Clarifications, drywell spray is secured when drywell pressure reaches 1.68 psig.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control Attached w/ Revision #

PPM 5.0.10, Flowchart Training Manual See Comments / Reference OI-15, EOP and EAL Clarifications Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8333 - Given a list, identify the statement that describes the two possible results of continuing to spray the wetwell when wetwell pressure is below 0 psig.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of termination point for drywell spray.

10 CFR Part 55 Content: 55.41 8 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-22 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-22 Comments /

Reference:

PPM 5.0.10 Rev: Major: Maj Minor: Min Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-22 Comments /

Reference:

OI-15 Rev: Major: 032 Minor: N/A Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-23 Examination Outline Cross-reference: 23 Revision: 0 Date: 06/30/20 Tier: 1 Group: 1 K/A Number: 295012 AK3.01 Level of Difficulty: 2 RO Importance Rating: 3.5 K/A

Description:

Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL TEMPERATURE: Increased drywell cooling A transient occurs that results in the following plant conditions:

The reactor is scrammed.

CB-S3 failed to automatically close

Drywell temperature slowly rising

Drywell pressure slowly rising.

The CRS has entered PPM 5.2.1, Primary Containment Control.

Which of the following actions should be taken now to lower Drywell temperature and pressure?

A. Perform PPM 5.5.15 Emergency Drywell Venting.

B. Manually close RCC-V-6 to divert all RCC cooling flow to the drywell loads.

C. Open the Weir Gate for the operating TSW Pump 200 turns.

D. Restart Drywell Cooling Fans.

Answer: D K/A Match:

Requires an understanding of system inter-relationships between TSW, RCC and Drywell Cooling Loads and the effects on Drywell temperature and pressure.

Explanation:

A. Incorrect. Plausible While PPM 5.5.15 would certainly lower Drywell pressure, however PPM 5.5.14 provides the direction necessary to vent the drywell during emergencies when the integrity of the Primary Containment is threatened. The stem of the question states we are in Mode 1 with drywell temperature and pressure slowly rising, we are not in EOPs and use of this procedure is not warranted for this situation. SOP-CN-CONT-VENT could be performed with these given conditions, PPM 5.514 is not performed until step P-14 of PPM 5.2.1 after emergency depressurization and WW Pressure cannot be maintained below PCPL. It is not performed from 100% power.

B. Incorrect. Plausible if the operator believes that RCC flow is not restored due to the failure of CB-S3. The failure of CB-S3 to automatically close will result in TR-B re-energizing SM-8. RCC pump Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-23 breakers do not trip on a loss of bus voltage (either temporary or complete loss) When SM-8 is re-energized RCC will automatically be restored and no actions are required to manually close RCC-V-6. (RCC-V-6 Noun Name not provided in the distractor; this valve is a required element of the must know RCC drawing for licenses candidates)

C. Incorrect. Step DT-1 bases states that the normal method of drywell temperature reduction is attempted in advance of initiating more complex actions to terminate increasing drywell temperature. Opening the Weir Gate for the operating TSW pump will increase the amount of cooler water from TMU to mix with the warmer CW basin water. TSW temperature will lower.

TSW cools the RCC Heat exchangers, a lower TSW temperature will result in more cooling in the RCC heat exchanger and lower RCC temperatures. The affect on RCC cooling is small (compared to restarting drywell cooling fans) and it not a standard action that is taken for this situation.

D. Correct. Failure of S-3 to automatically close following the scram results in power being temporarily lost to SM-8. When power is automatically restored to SM-8 by TR-B, there is no automatic function for restart of the drywell cooling fans. This is a manual action performed by the operator per the direction of the CRS via step DT-1 of PPM 5.2.1 Primary containment control Technical Reference(s)

PPM 5.0.10 Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-23 Proposed references to be provided during examination: N/A Learning Objective: a. 8311 Given a list, identify the statement that describes the purpose of attempting to maintain drywell temperature below 135 degrees F. (PPM 5.2.1)

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Needs solid understanding of system inter-relationships of RCC / TSW / Drywell cooling loads and correct situational mitigation strategies of PPM 5.2.1 based on synthesis of plant conditions.

10 CFR Part 55 Content: 55.41 5 Comments /

Reference:

PPM 5.0.10 Rev: Major: 14 Minor: 4 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-23 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-24 Examination Outline Cross-reference: 24 Revision: 0 Date: 8/7/20 Tier: 1 Group: 1 K/A Number: 295014.AK2.03 Level of Difficulty: 3 RO Importance Rating: 3.3 K/A

Description:

Knowledge of the interrelations between INADVERTENT REACTIVITY ADDITION and the following Fuel temperature CGS is in Mode 1.

The plant has just returned to full power following a refueling outage.

Reactor power is 100%.

Feedwater Heater 5A trips.

The CRS enters ABN-POWER, Unplanned Reactor Power Change.

How does this affect the core?

Power peaks at the core (1) . This effect causes flow instabilities and is more significant with (2) core flow.

A. (1) bottom (2) higher B. (1) bottom (2) lower C. (1) top (2) higher D. (1) top (2) lower Answer: B K/A Match:

Requires knowledge on how an inadvertent reactivity addition (loss of feedwater heater) affects core stability/fuel temperature.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-24 Explanation:

A. Incorrect. Plausible since power peaks at the bottom of the core. However, this is more significant at lower core flows.

B. Correct. In accordance with ABN-POWER, a trip of a feedwater heater #5 or #6 causes power to peak at the core bottom. In accordance with ABN-CORE, this effect is more pronounced with low core flow (i.e. single loop)

C. Incorrect. Plausible if it is believed that this effect is more pronounced with higher core flow (higher power). See explanation B.

D. Incorrect. Plausible since lower flow will make instabilities more pronounced. However, power will peak at the core bottom.

Technical Reference(s)

ABN-POWER, Unplanned Reactor Power Change Attached w/ Revision #

ABN-CORE, Unplanned Core Operating Conditions See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 6743 - With the procedures available, describe the reason for the following step associated with ABN-POWER: [ABN-POWER]: The requirement to reduce RRC flow to LE 80 Mlbm/hr, and insert rods to maintain below the 100% rod line when feedwater inlet temperature experiences an unplanned drop of GE 6°F.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of conditions that make power peaking and flow instabilities more severe.

10 CFR Part 55 Content: 55.41 1 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-24 Comments /

Reference:

ABN-POWER Rev: Major: 016 Minor: 003 Comments /

Reference:

ABN-CORE Rev: Major: Maj Minor: Min Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-25 Examination Outline Cross-reference: 25 Revision: 0 Date: 5/12/20 Tier: 1 Group: 2 K/A Number: 295022.AK1.01 Level of Difficulty: 3 RO Importance Rating: 3.3 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to LOSS OF CRD PUMPS: Reactor pressure vs. rod insertion capability The plant was operating at 100% power when the following occurs:

Time Event 1159 The operating Control Rod Drive (CRD) pump, CRD-P-1A, trips.

1200 Charging Header Pressure (CRD-PI-15) is 938 psig, down fast.

1203 Control Rod Scram Accumulator CRD-HCU-3027 is declared inoperable due to low accumulator pressure.

1205 Control Rod Scram Accumulator CRD-HCU-4635 is declared inoperable due to low accumulator pressure.

1230 Charging Header Pressure (CRD-PI-15) is 450 psig, down slow.

Based on this timeline, what is the time that a reactor scram is required to be initiated to meet technical specification requirements?

A. 1220 B. 1223 C. 1225 D. 1230 Answer: C K/A Match:

Requires knowledge of the operational requirements in Technical Specifications due to slow rod scram times on a loss of CRD pumps based on reactor pressure.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-25 Explanation:

A. Incorrect. Plausible if it is believed that a reactor scram must be initiated 20 minutes after the CRD pump trips with two inoperable scram accumulators. However, the 20 minute time limit starts when the second scram accumulator becomes inoperable with steam dome pressure 900 psig.

B. Incorrect. Plausible since a reactor scram is required 20 minutes after 2 control rod scram accumulators become inoperable with CRD d/p below 940 psig. However, the second control rod scram accumulator does not become inoperable until 1205.

Therefore, a scram is not required at 1223.

C. Correct. Limiting Condition for Operation (LCO) 3.1.5, Control Rod Scram Accumulators, Condition B states that with Two or more control rod scram accumulators inoperable with reactor steam dome pressure 900 psig, operators are required to Restore charging water header pressure to 940 psig with a required completion time (B.1) of 20 minutes from discovery of Condition B concurrent with charging water header pressure < 940 psig. For the conditions given, the clock for completing the required actions of Condition B starts at 1205, when the second scram accumulator becomes inoperable. Condition D of the LCO requires placing the Reactor Mod Switch in SHUTDOWN immediately if the completion time for required action B.1 is not met.

Therefore, the earliest time when a scram is required is 20 minutes from entering Condition B, or 1225.

D. Incorrect. Plausible since a reactor scram is required if steam dome pressure is < 900 psig and a single scram accumulator is inoperable with charging water pressure < 940 psig. However, the question stem states that at 1230, charging water pressure is < 900 psig, NOT steam dome pressure. With reactor power at 100%, steam dome pressure is well above 900 psig.

Technical Reference(s)

Technical Specifications Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11768 - Discuss when the reactor must be manually scrammed due to Control Rod Drive Hydraulic system malfunctions Question Source: #: 64695 (Sys ID)

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-25 Question Cognitive Level:

Justification for Cognitive Level Examinee must synthesize a knowledge of the < 1hour actions required by LCO 3.1.5 along with an understanding of when the plant conditions given in the question stem meet the requirements to take Tech Spec actions.

10 CFR Part 55 Content: 55.41 10 Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-25 Comments /

Reference:

Technical Specification Ammendment: 254 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-25 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-26 Examination Outline Cross-reference: 26 Revision: 0 Date: 6/9/20 Tier: 1 Group: 2 K/A Number: 295034.EA1.01 Level of Difficulty: 2 RO Importance Rating: 3.8 K/A

Description:

Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Area radiation monitoring system.

CGS is in Mode 1.

An event occurs that causes a reactor scram.

The following conditions exist:

Drywell pressure is 1.59 psig and up slow.

Suppression Pool level is -1.25 inches and down slow.

Control rod 30-31 is at position 48. All other control rods are at position 00.

ARM-RIS-4 East CRD Area is in alarm at 2300 mr/hr.

Reactor Building differential pressure is -.05 inches of water.

Reactor Building Exhaust Plenum is 12 mr/hr Based only on these conditions, which procedure should be used to address this event?

Enter PPM 5.1.1, RPV Control...

A. ONLY.

B. and PPM 5.1.2, RPV Control - ATWS.

C. and PPM 5.2.1, Primary Containment Control.

D. and PPM 5.3.1, Secondary Containment Control.

Answer: D K/A Match:

Requires knowledge of EOP entry conditions with respect to Secondary Containment Ventilation radiation levels.

Explanation:

A. Incorrect. Plausible since PPM 5.1.1. is entered on high drywell pressure and drywell pressure is above normal. However, PPM 5.1.1. is not entered until drywell pressure reaches 1.68 psig.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-26 B. Incorrect. since one control rod is at 48 inches. However, to enter PPM 5.1.2, at least two control rods must be GE Notch 02. Additionally, PPM 5.1.2 is only entered from PPM 5.1.1, which would not be entered for the conditions given in the question stem.

C. Incorrect. Plausible since entry into PPM 5.2.1 is required when Wetwell lowering. However, PPM 5.2.1 is not entered until Wetwell level is LT -2 inches.

D. Correct. One entry condition for PPM 5.3.1 is RB area radiation level above alarm level. The question stem states that ARM-RIS-4 is in alarm, which is an entry condition for PPM 5.3.1.

Technical Reference(s)

PPM 5.3.1, Secondary Containment Control Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8017 - Given plant conditions, recognize an EOP entry condition(s) and enter the appropriate flow chart.

Question Source: #: LO01149

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must synthesize information in question stem with a knowledge of entry conditions for multiple EOPs.

10 CFR Part 55 Content: 55.41 12 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-26 Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor: Min Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-27 Examination Outline Cross-reference: 27 Revision: 0 Date: 8/10/20 Tier: 1 Group: 2 K/A Number: 500000.2.4.50 Level of Difficulty: 3 RO Importance Rating: 4.2 K/A

Description:

High primary containment hydrogen: Ability to verify system alarm setpoints and operate controls identified in the alarm response manual.

CGS is in Mode 1.

The following annunciator is in alarm:

814.J2.2-2: CONTAINMENT ATMOSPHERE HYDROGEN LEVEL HIGH Primary Containment H2 indications are:

(CMS-CP-1301) (CMS-CP-1401)

R-1 In accordance with the Alarm Response Procedure (ARP) for 814.J2.2-2, what is the first action that should be performed?

A. Place the Hydrogen Water Chemistry Enable switch to SHUTDOWN (H13-P840) in accordance with PPM 5.2.1, Primary Containment Control.

B. Verify channel B operability by aligning channel A (CMS-CP-1301) to match the area being sampled on channel B (CMS-CP-1401).

C. Vent primary containment in accordance with PPM 5.5.21, Emergency Drywell Venting With High Hydrogen and Oxygen Concentrations.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-27 D. Perform a hydrogen monitor channel calibration in accordance with PPM 5.8.1, Post-LOCA Hydrogen/Oxygen Monitoring.

Answer: B K/A Match:

Requires knowledge of alarm procedure and equipment that needs to be operated for a high containment hydrogen alarm.

Explanation:

A. Incorrect. Plausible since the alarm setpoint for 814.J2.2-2 is 3.56%, which is entry criteria for PPM 5.2.1. However, the ARP directs actions to validate that the reading is correct, especially since the other monitor reads 0%.

B. Correct. In accordance with the ARP, the first action is to validate the hydrogen level by comparing the alarming channel to the other channel and aligning both channels to sample from the same location. Operators should wait for at least 15 minutes after shifting sample points to ensure a valid reading.

C. Incorrect. Plausible since this action would be performed in accordance with PPM 5.2.1, step H-3, if high hydrogen concentration is confirmed and PPM 5.2.1 is entered. However, for the conditions given in the question stem, the crew would not enter PPM 5.2.1 without first confirming hydrogen concentration.

D. Incorrect. Plausible since this calibration is mentioned in the ARP. However, it is performed post-LOCA to correct hydrogen monitor readings for elevated radiation levels. It is not required to be performed for the conditions given in the question stem.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control 4.814.J2.2-2 Containment Atmosphere Hydrogen Level High Alarm Attached w/ Revision #

Response Procedure See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8417 - Given PPM 5.2.1, "Primary Containment Control", identify the two methods of determining hydrogen and oxygen concentrations stated on the flow chart. (PPM 5.2.1)

Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-27 Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of actions for a high containment hydrogen concentration.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-27 Comments /

Reference:

4.814.J2.2-2 Rev: Major: 14 Minor: N/A Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-27 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-28 Examination Outline Cross-reference: 28 Revision: 0 Date: 8/10/20 Tier: 2 Group: 1 K/A Number: 203000.K6.01 Level of Difficulty: 2 RO Importance Rating: 3.6 K/A

Description:

Knowledge of the effect that a loss or malfunction of the following will have on the RHR/LPCI: INJECTION MODE: A.C. electrical power CGS is in Mode 1.

Startup Transformer, TR-S is out of service.

A LOCA occurs. An automatic scram occurs on high drywell pressure.

One minute later, a loss of the Normal Transformer (TR-N) occurs.

If SM-7 and SM-8 are reenergized at time t=0, what is the earliest time that RHR-P-2A will be running?

RHR-P-2A will be running at t=

A. 1 second.

B. 6 seconds C. 10 seconds D. 20 seconds Answer: B K/A Match:

Requires knowledge of how a loss of A.C. power supply affects RHR pump operation.

Explanation:

A. Incorrect. Plausible since RHR-P-2C will be running at 1 second. However, RHR-P-2A and RHR-P-2B have a 5 second delay prior to starting.

B. Correct. RHR-P-2A will start 5 seconds after SM-7 is reenergized from either TR-B or DG-1.

C. Incorrect. Plausible since this is the earliest time that RHR-P-2C would be running if TR-S picked up SM-1 and SM-7. However, TR-S is not available and there is a 5 second time delay for starting RHR-P-2A on TR-B.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-28 D. Incorrect. Plausible since this is the earliest time that RHR-P-2A would be running if TR-S picked up SM-1 and SM-7 (19.4 second time delay). However, since TR-B picked up SM-7, there is a 5 second time delay for starting RHR-P-2A.

Technical Reference(s)

SD000198, Residual Heat Removal System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5779 - Describe the expected system response, for any routine lineup, when the initiation logic for the LPCI mode of the RHR system is satisfied.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of start timers for RHR pumps.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-28 Comments /

Reference:

SD000198 Rev: Major: 17 Minor: 1 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-29 Examination Outline Cross-reference: 29 Revision: 0 Date: 5/6/20 Tier: 2 Group: 1 K/A Number: 205000.K4.01 Level of Difficulty: 2 RO Importance Rating: 3.4 K/A

Description:

Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following: High temperature isolation Which of the following signals will cause an isolation of NS4 Group 6, RHR (Shutdown Cooling Mode)?

(1) RPV Level 3 (LE +13 inches)

(2) Drywell high pressure ( GE 1.68 psig)

(3) RHR Area T high (GE 55°F)

A. (1) ONLY B. (2) ONLY C. (1) and (3) ONLY D. (2) and (3) ONLY Answer: C K/A Match:

Requires knowledge of the interlock that isolates RHR Shutdown Cooling on a High RHR Room Temperature.

Explanation:

A. Incorrect. Plausible since (1) will isolate Group 6. However, it is not the only signal listed that will isolate Group 6.

B. Incorrect. Plausible since drywell high pressure will isolate Group 3, 4, and 5. However, it is not an isolation signal for Group 6.

C. Correct. Group 6 is isolated by the following signals:

RPV Level 3 (LE +13 inches)

Reactor Pressure (GE 125 psig)

RHR Area Temp (GE 130°F to 150°F)

RHR Area T (GE 55°F)

Excess RHR SDC Flow (GE 6.28 psid)

D. Incorrect. Plausible since (3) will isolate Group 6. However, (2) is not an isolation signal for Group 6.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-29 Technical Reference(s)

SD000173, Nuclear Steam Supply Shutoff System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11811 - Describe the Residual Heat Removal System design features and/or interlocks which provide for the following: j. high temperature isolation in SDC mode.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the NS4 isolation circuitry for RHR in Shutdown Cooling mode.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-29 Comments /

Reference:

SD000173 Rev: Major: 14 Minor: 5 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Examination Outline Cross-reference: 30 Revision: 0 Date: 10/27/20 Tier: 2 Group: 1 K/A Number: 209001.K5.05 Level of Difficulty: 2 RO Importance Rating: 2.5 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM : System venting CGS is in Mode 1.

An event occurs which causes the following:

What action should be performed?

A. Manually start LPCS-P-1. Run the pump until the Keep Fill system is restored.

B. Prevent LPCS-P-1 from automatically starting by removing control power fuses.

C. Pressurize the LPCS system using the Condensate Storage and Transfer system.

Page 1 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 D. Fill the LPCS system with RHR-P-2A in accordance with ABN-LPCS-DEPRESS, LPCS Recovery Following Depressurization from Keep Fill Failure.

Answer: A K/A Match:

Requires knowledge of actions required when system used to keep LPCS system filled and vented fails.

Explanation:

A. Correct. In accordance with Alarm Response Procedure (ARP) 4.601.A3.6-3, LOW PRESSURE CORE SPRAY OUT OF SERVICE, Page 7 for LPCS BYPASS AND INOPERABLE STATUS PANEL (BISP) light for Loss of Power for LPCS-P-2, if the LPCS Pump Discharge Pressure Hi/Low Annunciator (4.601.A3.5-3) is not illuminated, LPCS-P-1 should be started and placed in Suppression Pool Mixing.

B. Incorrect. Plausible since the ARP for the LPCS Pump Discharge Pressure Hi/Low Annunciator (4.601.A3.5-3) directs removing the control power fuses if the alarm is annunciating. However, for the conditions given, this action is not required.

C. Incorrect. Plausible since the Condensate Storage and Transfer system is used to fill LPCS following maintenance. However, precaution 4.3 in SOP-LPCS-FILL, LPCS Fill and Vent, prohibits using this system to keep LPCS pressurized.

D. Plausible since ABN-LPCS-DEPRESS allows using RHR-A to pressurize LPCS. However, this procedure is entered after LPCS is depressurized due to keep-fill system failure (i.e. the LPCS low pressure alarm is in).

Technical Reference(s) 4.601.A3, Alarm Response Procedure for panel 4.601.A3 SOP-LPCS-FILL, LPCS Fill and Vent Attached w/ Revision #

ABN-LPCS-DEPRESS, LPCS Recovery Following Depressuration See Comments / Reference from Keep Fill Failure Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11590 - Describe the effect that a loss of malfunction of the following will have on the Low Pressure Core Spray System: c. Keep fill system Page 2 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of actions required when the LPCS keep-fill system fails.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

4.601.A3 Rev: Major: 028 Minor: N/A Page 3 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Page 4 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Page 5 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Comments /

Reference:

SOP-LPCS-FILL Rev: Major: 15 Minor: N/A Page 6 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-30 Comments /

Reference:

ABN-LPC-DEPRESS Rev: Major: 002 Minor: 001 Page 7 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 Examination Outline Cross-reference: 31 Revision: 0 Date: 5/13/20 Tier: 2 Group: 1 K/A Number: 209002.K4.02 Level of Difficulty: 2 RO Importance Rating: 3.4 K/A

Description:

Knowledge of HIGH PRESSURE CORE SPRAY SYSTEM (HPCS) design feature(s) and/or interlocks which provide for the following: Prevents over filling reactor vessel CGS is in Mode 1

A LOCA occurs

HPCS initiates on low RPV level.

Drywell Pressure is 1.2 psig, up slow.

RPV Level is +30 inches, up fast.

The HPCS RPV Injection Valve, HPCS-V-4, automatically closes when RPV level reaches Level-8.

RPV level is 55 inches, down slow.

Which of the following conditions, when considered separately, will cause HPCS-V-4 to automatically re-open?

A. RPV level goes below the Level-8 setpoint (+54.5 inches).

B. Drywell Pressure goes above 1.68 psig.

C. RPV level goes below the Level-2 setpoint (-50 inches).

D. Wetwell level goes above +5 inches.

Answer: C K/A Match:

Requires knowledge of the automatic operating setpoint of HPCS-V-4, which prevents over filling the RPV while maintaining RPV level.

Explanation:

A. Incorrect. Plausible since HPCS-V-4 closes when level is > Level 8 and it may be assumed that HPCS-V-4 toggles closed and open around the Level 8 setpoint. However, once HPCS is initiated and RPV level goes above Level 8. HPCS-V-4 will not automatically open until RPV level goes below the Level 2 setpoint.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 B. Incorrect. Plausible since Drywell pressure is an initiation signal for HPCS and will normally cause HPCS-V-4 to open. However, once HPCS is initiated and RPV level goes above Level 8.

HPCS-V-4 will not automatically open until RPV level goes below the Level 2 setpoint.

C. Correct. After HPCS initiates, HPCS-V-4 will automatically close when RPV level reaches Level 8

(+54.5 inches). HPCS-V-4 will remain closed until RPV level goes below Level 2 (-50 inches) to prevent over filling the RPV.

D. Incorrect. Plausible since the HPCS suction source automatically swaps to the wetwell from the condensate storage tank when wetwell level goes above +5 inches, and it may be assumed that HPCS-V-4 opens to help prevent wetwell level from continuing to rise. However, once HPCS is initiated and RPV level goes above Level 8. HPCS-V-4 will not automatically open until RPV level goes below the Level 2 setpoint.

Technical Reference(s)

SD000174 - HPCS Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: Learning Objective.

Question Source: #: LX00777

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must synthesize knowledge of the automatic setpoints of HPCS-V-4 with the plant conditions given in the question stem.

10 CFR Part 55 Content: 55.41 8 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 Comments /

Reference:

SD000174 Rev: Major: 16 Minor: 3 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-31 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-32 Examination Outline Cross-reference: 32 Revision: 0 Date: 5/13/20 Tier: 2 Group: 1 K/A Number: 211000.A4.03 Level of Difficulty: 2 RO Importance Rating: 4.1 K/A

Description:

Ability to manually operate and/or monitor in the control room: Explosive valves firing circuit status CGS is in Mode 1.

In response to a LOCA the Standby Liquid Control (SLC) system is in the following condition:

(1) What is the status of the Squib Discharge Valves?

(2) What is the status of the SLC Injection Pumps?

A. (1) SLC-V-4A is OPEN.

(2) ONLY SLC-P-1A is injecting.

B. (1) SLC-V-4A is OPEN.

(2) BOTH SLC-P-1A and SLC-P-1B are injecting.

C. (1) SLC-V-4B is OPEN.

(2) ONLY SLC-P-1A is injecting.

D. (1) SLC-V-4B is OPEN.

(2) BOTH SLC-P-1A and SLC-P-1B are injecting.

Answer: D Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-32 K/A Match:

Requires knowledge of the indications that a SLC Squib Injection Valve is open/closed.

Explanation:

A. Incorrect. Plausible if it is believed that the squib valve Circuit Ready light is ON when the squib has fired and that the squib injection valve must be open for the associated injection pump to inject. However, the SLC-V-4A "Circuit Ready" light ON indicates that the squib for SLC-V-4A did not fire and the valve is closed. Additionally, there is a cross connect on the pump discharge piping and both pumps will inject through the open squib valve (SLC-V-4B).

B. Incorrect. Plausible since there is a cross connect on the pump discharge piping and both pumps will inject through the open squib valve (SLC-V-4B). However, the SLC-V-4A "Circuit Ready" light ON indicates that the squib for SLC-V-4A did not fire and the valve is closed.

C. Incorrect. Plausible since SLC-V-4B is open. However, there is a cross connect on the pump discharge piping and both pumps will inject through the open squib valve (SLC-V-4B).

D. Correct. When SLC System A and B switches are in OPER, SLC-V-1A & B are full open, and SLC-V-31 is closed, SLC-P-1A & B will start. The SLC-V-4B "Circuit Ready" light OFF indicates that the squib for SLC-V-4B fired and the valve is open. There is a cross connect on the pump discharge piping and both pumps will inject through the open squib valve (SLC-V-4B).

Technical Reference(s)

SD000172 - Standby Liquid Control Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5923 - Describe the operation of the following SLC System components:

a. Pumps, d. Explosive Valves Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-32 Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing system information given in the stem with a knowledge of SLC squib valve indicatons and SLC pump configuration.

10 CFR Part 55 Content: 55.41 8 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-32 Comments /

Reference:

SD000172 Rev: Major: 13 Minor: 1 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-33 Examination Outline Cross-reference: 33 Revision: 0 Date: 8/5/20 Tier: 2 Group: 1 K/A Number: 212000.A4.01 Level of Difficulty: 2 RO Importance Rating: 4.6 K/A

Description:

Ability to manually operate and/or monitor in the control room: Provide manual SCRAM signal(s)

CGS is in Mode 1.

The crew needs to insert a manual reactor scram.

Which of the following combinations of Reactor Scram pushbuttons on H13-P603, when depressed simultaneously, will initiate a reactor scram?

(1) Logic A1 and Logic A2 (2) Logic A1 and Logic B2 (3) Logic A2 and Logic B2 A. (1) ONLY B. (2) ONLY C. (1) and (3) ONLY D. (2) and (3) ONLY Answer: D K/A Match:

Requires knowledge of the method of manually scramming the reactor using Reactor Scram pushbuttons.

Explanation:

A. Incorrect. Plausible if it is believed that both pushbuttons from the same Trip System (A or B) must be depressed to initiate a scram. However, this will only cause a half-scram on Trip System A.

B. Incorrect. Plausible since this combination will cause a scram. Plausible if it is believed that the pushbuttons must be from different Trip Channels (1 or 2) to cause a scram. However, it is not the only combination listed that will cause a scram.

C. Incorrect. Plausible since (3) will cause a scram. However, (1) will only cause a half-scram.

D. Correct. Each RPS is divided into two Trip Systems, A and B. Each Trip System is comprised of two redundant Trip Channels, 1 and 2. To initiate a scram, one Trip Channel from each Trip Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-33 System must be activated. Depressing any combination of A1 or A2 and B1 or B2 causes a full scram. Therefore, (2) and (3) will initiate a reactor scram.

Technical Reference(s)

SD000161, Reactor Protection System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11681 - Describe the function, purpose, and design features of the major RPS components: l. Manual Scram pushbuttons.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the function of the Reactor Scram pushbuttons.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-33 Comments /

Reference:

SD000161 Rev: Major: 17 Minor: 5 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-34 Examination Outline Cross-reference: 34 Revision: 0 Date: 5/28/20 Tier: 2 Group: 1 K/A Number: 215003.K1.05 Level of Difficulty: 2 RO Importance Rating: 3.3 K/A

Description:

Knowledge of the physical connections and/or cause-effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following: Display control system Intermediate range monitor (IRM) channel "E" is currently on range 7 reading 10/40 scale.

How does this indication change when the IRM scale is changed?

If channel "E" is placed in range (1) , it will be reading (2) scale.

A. (1) 6 (2) 1/125 B. (1) 6 (2) 10/125 C. (1) 8 (2) 10/125 D. (1) 8 (2) 100/125 Answer: C K/A Match:

Requires understanding of changes in IRM display when the range switch is operated.

Explanation:

A. Incorrect. Plausible if it is believed that the IRM indication would lower when going on a lower scale. However, taking IRM E to range scale would increase the displayed value to 100/125 scale.

B. Incorrect. Plausible if it is believed that the odd ranges are expanded versions of the range below them. However, they are an expanded version of the scale above. Going from an odd scale to the even scale below will cause output to increase by a factor of 10.

C. Correct. IRM odd ranges are an expanded version of the even scale above. The even scale is the same as the 0-40 portion of the odd scale above. Therefore when the IRM reads 10/40scale on range 6, taking the range switch to the next odd range above (range 8) will cause the display to read 10/125 scale.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-34 D. Incorrect. Plausible if it is believed that taking the IRM from an odd range to the next higher even range will cause output to increase by a factor of 10. However, taking the range switch from an odd range to the next lower even range will cause output to increase by a factor of 10. Going to the next higher even range will give an output that is the same, but on a larger scale. 10/40 scale on range 7 will go to 10/125 scale on range 8.

Technical Reference(s)

SD000138, IRM System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5453 - Describe the relationship of readings between IRM even/odd ranges, even/even ranges, and odd/odd ranges.

Question Source: #: LO00650

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the operation of the IRM range switches.

10 CFR Part 55 Content: 55.41 6 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-34 Comments /

Reference:

SD000138 Rev: Major: 10 Minor: 2 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-35 Examination Outline Cross-reference: 35 Revision: 0 Date: 5/27/20 Tier: 2 Group: 1 K/A Number: 215004.K2.01 Level of Difficulty: 2 RO Importance Rating: 2.6 K/A

Description:

Knowledge of electrical power supplies to the following: SRM channels/detectors What is the power supply to SRM Channel C?

A. DP-SO-A B. DP-SO-B C. RPS-A D. RPS-B Answer: A K/A Match:

Requires knowledge of SRM power supplies.

Explanation:

A. Correct. In accordance with the Source Range Monitor System Description (SD000132), the power supply to Division 1 SRM Channels A and C is DP-SO-A B. Incorrect. Plausible if it is believed that SRM Channel C is Division 2. However, it is Division 1 C. Incorrect. Plausible since Division 1 Power Range Nuclear Monitoring (PRNM) is powered from RPS-A. However, Division 1 SRMs are powered from DP-SO-A.

D. Incorrect. Plausible since RPS-B is the power supply to one Rod Block Monitor interface panel and Division 2 PRNM. However, the power supply to SRM Channel C is DP-SO-A.

Technical Reference(s)

SD000132, Source Range Monitors Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 12000 - State the power supplies to the following Source Range Monitoring components:

Page 1 of 2

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-35

a. SRM channels/detectors

b. Detector drive modules

c. Detector drive module control Question Source: #: LO03583

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires a knowledge of the power supply to SRM Channel A.

10 CFR Part 55 Content: 55.41 7 Comments /

Reference:

SD000132 Rev: Major: 12 Minor: 2 Page 2 of 2

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-36 Examination Outline Cross-reference: 36 Revision: 0 Date: 5/28/20 Tier: 2 Group: 1 K/A Number: 215005.A3.07 Level of Difficulty: 4 RO Importance Rating: 3.8 K/A

Description:

Ability to monitor automatic operations of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM including: RPS status CGS is in Mode 1.

A transient has resulted in the following:

Core Flow: 50%

APRM-1 Reactor Power: 86%

APRM-2 Reactor Power: 87%

APRM-3 Reactor Power: 88%

APRM-4 Reactor Power: 95%

What is the effect of these APRM outputs?

A rod block (1) present. A half-scram (2) present.

A. (1) is (2) is B. (1) is (2) is not C. (1) is not (2) is D. (1) is not (2) is not Answer: B K/A Match:

Requires understanding of the relationship between APRM output and RPS status.

Explanation:

A. Incorrect. Plausible since a rod block is present for the given conditions. However, a half-scram does not occur for any combination of above the scram setpoints. See explanation B below.

B. Correct. Each ARPM channel calculates setpoints based on current core flow as follows:

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-36

Rod Block setpoint = 0.62(Core Flow %) + 57.1% = 88.1% for the conditions given.

Scram setpoint = 0.62(Core Flow %) + 60.9% = 91.9% for the conditions given.

Therefore, APRM-4 has exceeded both the rod block and scram setpoint. No other channel has exceeded a setpoint.

A Rod Block can be generated from any single APRM chassis. Rod Blocks DO NOT use 2 out-of-4 logic. The Voter need two similar APRM trips to actuate (e.g. two APRM Upscale Trips or two OPRM Trips). The logic is normally 2 out of 4 unless an APRM is bypassed or inoperative, then the logic will be 2 out of 3. The output of the Voter is arranged such that there are two relay outputs (X and Y) that de-energize to actuate the respective RPS logic. Voter 1 and 3 input into the RPS Channels A1 and A2. Voter 2 and 4 input into the RPS Channels B1 and B2. Since all Voters receive the input from all operable APRMs, when at least two APRM trip signal are generated, a full reactor scram will occur. A half-scram condition will only exist when a single Voter is deenergized or during testing.

C. Incorrect. Plausible if it is believed that the rod block follows two of four logic while the RPS will generate a scram since all APRM Voters receive input from all APRM channels. However, only a rod block will be present. See explanation B above.

D. Incorrect. Plausible since a half-scram is not generated. However, a rod block is generated. See explanation B above.

Technical Reference(s)

SD001819, PRNM System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 13711 - Describe PRNM system automatic initiation signals, operation, controls, automatic functions, alarms, setpoints, and or interlocks.

Question Source: #: LO03335

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information given in the stem with an understanding of the logic used to create rod blocks, and a knowledge of the logic for scram signals.

10 CFR Part 55 Content: 55.41 6 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-36 Comments /

Reference:

SD001819 Rev: Major: 4 Minor: 0 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-37 Examination Outline Cross-reference: 37 Revision: 0 Date: 8/20/20 Tier: 2 Group: 1 K/A Number: 217000.A2.16 Level of Difficulty: 3 RO Importance Rating: 3.5 K/A

Description:

Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low condensate storage tank level.

CGS is in Mode 1.

A LOCA occurs.

The crew is taking actions in accordance with PPM 5.1.1, RPV Control.

RCIC is injecting into the RPV, maintaining RPV level.

All other high pressure injection sources are unavailable.

The following annunciator alarms:

4.601.A4.3-4, RCIC SUCTION SWITCHOVER CST LEVEL LOW RCIC suction valves are aligned as follows:

RCIC-V-10, Pump Suction From Condensate Storage Tank, is OPEN.

RCIC-V-31, Pump Suction From Suppression Pool, is CLOSED.

What action(s) should be taken?

A. Manually OPEN RCIC-V-31. RCIC-V-10 must be manually CLOSED after RCIC-V-31 is open.

B. Manually OPEN RCIC-V-31. RCIC-V-10 will automatically close after RCIC-V-31 is open.

C. Manually CLOSE RCIC-V-10. RCIC-V-31 must be manually OPENED after RCIC-V-10 is closed.

D. Manually CLOSE RCIC-V-10. RCIC-V-31 will automatically close after RCIC-V-10 is closed.

Answer: B Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-37 K/A Match:

Requires knowledge of actions required if RCIC suction valves fail to automatically swap on low condensate storage tank level.

Explanation:

A. Incorrect. Plausible since RCIC-V-31 failed to automatically open and must be manually opened.

However, RCIC-V-10 does not need to be manually closed. It will automatically close when RCIC-V-31 is fully open.

B. Correct. Annunciator 4.601.A4.3-4 in alarm means that CST level is at or below the RCIC suction auto-swap over level. RCIC-V-31 should have automatically opened. Since the valve is still closed, the operators should take actions in accordance with SOP-RCIC-SUCTION, RCIC Suction Transfer, which directs manually opening RCIC-V-31. RCIC-V-10 will automatically close when RCIC-V-31 is fully open.

C. Incorrect. Plausible since RCIC-V-10 should be closed for the conditions given in the question stem. However, RCIC-V-10 is not closed before RCIC-V-31 is fully open to prevent starving the suction of the RCIC pump.

D. Incorrect. Plausible since RCIC-V-10 should be closed for the conditions given in the question stem. Plausibility is enhanced since there is an interlock between RCIC-V-10 and RCIC-V-31.

However, RCIC-V-10 is not closed before RCIC-V-31 is fully open to prevent starving the suction of the RCIC pump. Additionally, RCIC-V-31 does not automatically open when RCIC-V-10 is closed.

Technical Reference(s) 4.601.A4.3-4, Alarm Response Procedure Attached w/ Revision #

SOP-RCIC-SUCTION, RCIC Suction Transfer See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5724 - Explain the interlock associated with: d. RCIC-V-10 and RCIC-V-31 Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-37 Justification for Cognitive Level Requires understanding of the interlock for RCIC pump suction valves.

10 CFR Part 55 Content: 55.41 6 Comments /

Reference:

4.601.A4.3-4 Rev: Major: 42 Minor: N/A Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-37 Comments /

Reference:

SOP-RCIC-SUCTION Rev: Major: 001 Minor: 001 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-38 Examination Outline Cross-reference: 38 Revision: 0 Date: 5/14/20 Tier: 2 Group: 1 K/A Number: 218000.K5.01 Level of Difficulty: 3 RO Importance Rating: 3.8 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION SYSTEM : ADS logic operation CGS is in Mode 1.

A small-break LOCA has occurred.

No high pressure injection sources are available.

All low pressure ECCS pumps are running with normal discharge pressures.

At 1200 the following plant conditions exist:

RPV level is lowering at constant rate of 5 inches/minute.

Assuming no operator action is taken, what is the earliest time that the ADS valves will open?

A. 1206 B. 1208 Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-38 C. 1221 D. 1223 Answer: D K/A Match:

Requires knowledge of the results of ADS logic actuation.

Explanation:

A. Incorrect. Plausible if it is believed that ADS initiates immediately upon an RPV Level 2 (-

50 inches). However, the ADS timer starts on RPV Level 1 (-129 inches) and actuates ADS 105 seconds later if RPV level remains below Level 1.

B. Incorrect. Plausible if it is believed that the ADS timer starts when RPV level goes below Level 2. However, the ADS timer starts on RPV Level 1 (-129 inches) and actuates ADS 105 seconds later if RPV level remains below Level 1.

C. Incorrect. Plausible if it is believed that ADS initiates immediately upon an RPV Level 1 (-

129 inches). However, the ADS timer starts on RPV Level 1 (-129 inches) and actuates ADS 105 seconds later if RPV level remains below Level 1.

D. Correct. The ADS timer starts on RPV Level 1 (-129 inches) and actuates ADS 105 seconds later if RPV level remains below Level 1. Therefore the ADS timer will start at 1220 and 48 seconds. The timer will time out 105 seconds later, or 1222 and 33 seconds, and the valves will be open by 1223.

Technical Reference(s)

SD000186 - Automatic Depressurization (ADS)

Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5071 - State the condition that will automatically initiate ADS. Include signals, setpoints, and time delays.

Question Source: #: LO01807

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: 2009 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-38 Question Cognitive Level:

Justification for Cognitive Level Examinee must evaluate the changes in plant conditions given in the question stem and synthesize this evaluation with a knowledge of ADS logic operation to determine when the ADS valves will open.

10 CFR Part 55 Content: 55.41 7 Comments /

Reference:

SD000186 Rev: Major: 12 Minor: 2 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-39 Examination Outline Cross-reference: 39 Revision: 0 Date: 8/19/20 Tier: 2 Group: 1 K/A Number: 223002.K3.20 Level of Difficulty: 3 RO Importance Rating: 3.3 K/A

Description:

Knowledge of the effect that a loss or malfunction of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following: Standby Gas Treatment System CGS is in Mode 1.

An NS4 Group 3 Outboard isolation signal is generated.

How is the Standby Gas Treatment (SGT) system affected?

1 minute after the fault, SGT Train A will be (1) and SGT Train B will be (2) .

A. (1) off (2) off B. (1) off (2) running C. (1) running (2) off D. (1) running (2) running Answer: C K/A Match:

Requires knowledge of the effect of a malfunction of NS4 on the Standby Gas Treatment system.

Explanation:

A. Incorrect. Plausible if it is believed that it takes both an Inboard and Outboard isolation signal to start SGT. However, SGT Train B starts on an Inboard signal while SGT Train A starts on an Outboard signal.

B. Incorrect. Plausible if it is believed that SGT Train A starts on an Inboard signal while SGT Train B starts on an Outboard signal. However, the actual logic is opposite of this.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-39 C. Correct. SGT Train B starts on an Inboard signal while SGT Train A starts on an Outboard signal.

D. Incorrect. Plausible if it is believed that both SGT trains start on either Outboard or Inboard isolation signal. However, SGT Train B starts on an Inboard signal while SGT Train A starts on an Outboard signal.

Technical Reference(s)

SD000173, NS4 System Description.

Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5596 - Describe the isolation logic used by the NS4 system for MSIV isolation and Group 3 and 4.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires understanding of NS4 isolation logic.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-39 Comments /

Reference:

SD000173 Rev: Major: 14 Minor: 5 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-40 Examination Outline Cross-reference: 40 Revision: 0 Date: 5/28/20 Tier: 2 Group: 1 K/A Number: 239002.A1.07 Level of Difficulty: 2 RO Importance Rating: 2.9 K/A

Description:

Ability to predict and/or monitor changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including: Turbine load CGS is in Mode 1.

A Safety Relief Valve (SRV) inadvertently opens.

With no operator actions, what is the condition of the reactor plant 9 minutes after the SRV opens when compared to conditions prior to the SRV opening?

RPV level will be (1) and Main Generator output will be (2) .

A. (1) lower (2) lower B. (1) lower (2) higher C. (1) higher (2) lower D. (1) higher (2) higher Answer: A K/A Match:

Requires knowledge of the change in turbine load when an SRV is opened at power.

Explanation:

A. Correct. When the SRV opens, DEH responds by closing the governor valves to restore the Main Steam equalizing header pressure to the DEH setpoint. This results in lower steam flow to the Main Turbine and lower Main Generator output (approximately 70 MWe). The steam passing through the open SRV is not seen by the steam flow instrumentation. The Feedwater Level Control system detects a feed flow to steam flow mismatch and lowers feed flow to match the new lower steam flow. This causes RPV level to be reduced to a lower steady state value.

B. Incorrect. Plausible since RPV level will be lower. However, Main Generator output will also be lower. See explanation A above.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-40 C. Incorrect. Plausible since Main Generator load will be lower. Plausibility is enhanced if it is believed that the open SRV will cause a steady state rise in total steam flow and RPV level rises due to swell. However, steady state RPV level will lower. See explanation A above.

D. Incorrect. Plausible if it is believed that the open SRV will cause a steady state rise in total steam flow and RPV level rises due to swell and if DEH raises steam flow to maintain Main Steam equalizing header pressure. However, both RPV level and Main Generator output will be lower.

See explanation A above.

Technical Reference(s)

ABN-SRV, Safety Relief Valve Opening Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11697 - Predict the impact on the following with an SRV open:

a. Tail pipe temperature

b. Reactor pressure, water level, power

c. Turbine load

d. Suppression pool water temperature

e. Indicated vs. actual steam flow

f. SRV stuck open Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must understand how the DEH system and the Feedwater Level Control system respond to an SRV opening at power.

10 CFR Part 55 Content: 55.41 5 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-40 Comments /

Reference:

ABN-SRV Rev: Major: 006 Minor: N/A Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-40 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-41 Examination Outline Cross-reference: 41 Revision: 0 Date: 8/19/20 Tier: 2 Group: 1 K/A Number: 259002.K1.09 Level of Difficulty: 3 RO Importance Rating: 2.9 K/A

Description:

Knowledge of the physical connections and/or cause-effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following: P sat/T sat (compensation)

Which RPV level instrument provides input to the Startup Level Controller (RFW-LIC-620) to control RPV level during low power level operation?

The (1) level instrument which (2) compensated for density changes in the RPV.

A. (1) Narrow Range (2) is B. (1) Narrow Range (2) is not C. (1) Wide Range (2) is D. (1) Wide Range (2) is not Answer: B K/A Match:

Requires knowledge that level input to the Feedwater Level Control (FWLC) system is not density compensated.

Explanation:

A. Incorrect. Plausible since Narrow Range provides input to RFW-LIC-620. However, the Narrow Range is not density compensated.

B. Correct. The Narrow Range RPV level instrument provides input to RFW-LIC-620. This range is not density compensated.

C. Incorrect. Plausible since the Wide Range instrument would provide a wider input range for low power operation. However, the Narrow Range instrument provides input to RFW-LIC-620.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-41 D. Incorrect. Plausible since the Wide Range instrument would provide a wider input range for low power operation and the Wide Range is not density compensated. However, the Narrow Range instrument provides input to RFW-LIC-620.

Technical Reference(s)

SD000157, Feedwater Level Control (FWLC) System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5394 - Describe the function of each of the following controls and how they relate to each other: a. Startup Level Controller Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the input to the Startup Level Controller and which RPV level instruments are density compensated.

10 CFR Part 55 Content: 55.41 4 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-41 Comments /

Reference:

SD000157 Rev: Major: 19 Minor: 1 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-42 Examination Outline Cross-reference: 42 Revision: 0 Date: 7/29/20 Tier: 2 Group: 1 K/A Number: 261000.A1.04 Level of Difficulty: 2 RO Importance Rating: 3.0 K/A

Description:

Ability to predict and/or monitor changes in parameters associated with operating the STANDBY GAS TREATMENT SYSTEM controls including: Secondary containment differential pressure CGS is in Mode 1.

Operators are starting SGT Division 1 to control Reactor Building differential pressure.

How is the SGT system aligned to control Reactor Building pressure?

The SGT Differential Pressure Indicating Controllers (DPICs) are normally set in auto at (1) inch(es) wg to ensure all portions of secondary containment are maintained at a maximum pressure of (2) inch(es) wg.

A. (1) -0.8 (2) -0.25 B. (1) -0.8 (2) -0.8 C. (1) -1.7 (2) -0.25 D. (1) -1.7 (2) -0.8 Answer: C K/A Match:

Requires knowledge of SGT system setup to control secondary containment differential pressure within allowable values.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-42 Explanation:

A. Incorrect. Plausible since the SGT DPICs are aligned to maintain secondary containment d/p at or below -0.25 inch wg. However, the DPICs are normally set to -1.7 inches wg.

B. Incorrect. Plausible since Reactor Building Ventilation is set to -0.8 inch wg to control secondary containment d/p. However, when SGT is used to control d/p, the SGT DPICs are normally set to -

1.7 inches wg.

C. Correct. When controlling secondary containment d/p, SGT DPICs are normally in auto set to -1.7 inches wg to ensure that all portions of the secondary containment structure are at or less than -

0.25 inch wg.

D. Incorrect. Plausible since SGT DPICs are normally set to -1.7 inches wg. However, this ensures that secondary containment d/p will be at or less than -0.25 inch d/p.

Technical Reference(s)

SD000144, Standby Gas Treatment System Description Attached w/ Revision #

SD000139, Secondary Containment System Descriptiion See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5822 - State the Reactor Building pressure the SGT system is designed to maintain, as well as the pressure its DPIC is set to maintain and why it is at that setting.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge to the normal setting for SGT DPICs and the reason that this setting is selected.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-42 Comments /

Reference:

SD000144 Rev: Major: 16 Minor: 1 Comments /

Reference:

SD000139 Rev: Major: 13 Minor: 1 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-43 Examination Outline Cross-reference: 43 Revision: 0 Date: 8/6/20 Tier: 2 Group: 1 K/A Number: 262001.A2.10 Level of Difficulty: 2 RO Importance Rating: 2.9 K/A

Description:

Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Exceeding current limitations CGS is in Mode 1.

A reactor plant startup is in progress in accordance with PPM 3.1.2, Start Up Flow Chart.

Reactor power is 15% and stable.

The crew has just synchronized the Main Generator to the grid.

Ashe Breaker No. 1 (BPA BKR 4885) is closed.

Ashe Breaker No.2 (BPA BKR 4888) is open.

TR-S is powering SM-1, SM-2, and SM-3.

A fault at the Ashe Substation causes TR-S relay 50/51T-SN (230 kV Neutral Overcurrent) to trip.

What action(s) should be taken?

(1) Verify the Main Turbine is tripped.

(2) Trip DG-3 by depressing the Engine Emergency Stop pushbutton.

(3) Verify RCC-V-6, RCC Radwaste/Rx Bldg Supply, is closed.

A. (1) ONLY B. (2) ONLY C. (1) and (3) ONLY D. (2) and (3) ONLY Answer: A K/A Match:

Requires examinee to understand what happens with a trip of TR-S and what immediate actions need to be taken.

Page 1 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-43 Explanation:

A. Correct. Since one of the 500 kV generator breakers is closed (BPA BKR 4885) and breakers S-1, S-2, and S-3 are all closed, a trip of relay 50/51T-SN will cause a Unit Lockout trip via relays 86XUOA and 86X1UOA. This will cause a Main Turbine Trip.

B. Incorrect. Plausible since a trip of relay 50/51T-SN will cause all S breakers to trip. This will cause an undervoltage condition on SM-2 and SM-4. DG-3 will start and reenergize SM-4. If HPCS-P-2, DG3 SW pump, were not running, the immediate action in accordance with ABN-ELEC-SM2/SM4 is to stop DG3 by pressing the Emergency Stop pushbutton. However, there is no reason given for HPCS-P-2 not to start. Therefore, this immediate action is not required.

C. Incorrect. Plausible since immediate action (1) is required. Plausibility is enhanced since, with the fault at the Ashe Substation, tie breaker 3-8 will open. If Bus SM-8 remains deenergized, an immediate action in accordance with ABN-ELEC-SM3/SM8 is to verify that RCC-V-6 is closed if SM-8 is deenergized and cannot be reenergized by TR-B or DG2. However, although SM-8 initially loses power, it will be reenergized by TR-B.

D. Incorrect. For plausibility, see explanation B and C.

Technical Reference(s)

SD000182, AC Distribution System Description.

ABN-ELEC-SM2/SM4, SM-2, SM-4 and SL-21 Distribution System Attached w/ Revision #

Failures ABN-ELEC-SM-3/SM-8, SM-3, SM-8, SM-85, SM-82, SL-81, SL-83 & See Comments / Reference SL-31 Distribution System Page: 3 of 27 Failures ABN-TURBINE, Main Turbine Generator Trip Proposed references to be provided during examination: N/A Learning Objective: 5055 - Identify the plant response to a "Loss Of TR-S During Startup" including the electric plant configuration necessary to cause this trip to occur.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Page 2 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-43 Examinee must demonstrate an understanding for the conditions that cause a Loss of TR-S During Startup along with an understanding of the required actions for this condition.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

SD000182 Rev: Major: 21 Minor: 3 Page 3 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-43 Page 4 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-43 Comments /

Reference:

ABN-ELEC-SM2/SM4 Rev: Major: Maj Minor: Min Page 5 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-43 Comments /

Reference:

ABN-ELEC-SM-3/SM-8 Rev: Major: 22 Minor: N/A Page 6 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-43 Comments /

Reference:

ABN-TURBINE Rev: Major: 003 Minor: N/A Page 7 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-44 Examination Outline Cross-reference: 44 Revision: 0 Date: 10/27/20 Tier: 2 Group: 1 K/A Number: 262002.K6.02 Level of Difficulty: 2 RO Importance Rating: 2.8 K/A

Description:

Knowledge of the effect that a loss or malfunction of the following will have on the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) : D.C. electrical power.

CGS is in Mode 1.

A field operator reports the following conditions on Inverter IN-3:

(Light ON) (Light ON) (Light ON)

What caused these indications?

Loss of bus A. S1-1 B. S1-2 C. MC-7A D. MC-8A Answer: A K/A Match:

Requires knowledge of how inverters respond to a loss of D.C. electrical power.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-44 Explanation:

A. Correct. The normal DC supply to IN-3 is S1-1. If S1-1 is lost, the inverter static switch will provide power to the load via the alternate AC supply, MC-7A. The Alternate Source Supplying Load light will energize.

B. Incorrect. Plausible since S1-2 is the normal DC supply for IN-2. However, S1-1 is the normal supply for IN-3.

C. Incorrect. Plausible since MC-7A is the alternate AC source for IN-3. This answer would be correct if IN-3 were similar to IN-1, which uses a normal AC supply. However, IN-3 uses S1-1 as a normal supply. The loss of MC-7A would not cause the inverter to switch to the alternate source, and the ALTERNATE SOURCE SUPPLYING LOAD light would be off.

D. Incorrect. Plausible since MC-8A is the alternate AC source for IN-2. This answer would be correct if a examinee believed this to be the supply to IN-3 and IN-3 were similar to IN-1, which uses a normal AC supply. However, IN-3 uses S1-1 as a normal supply.

Technical Reference(s)

SD000194, UPS System Description Attached w/ Revision #

ABN-ELEC-125VDC, Plant 125 VDC Distribution System Failures See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5894 - List the three (3) things that will cause an automatic transfer of E-IN-2 and E-IN-3.

Question Source: #: LO01368

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to have knowledge of the power supplies of IN-3, to understand the difference between the operation of IN-3 and other inverters (IN-1, IN-5) and to interpret the meaning of the ALTERNATE SOURCE SUPPLYING LOAD light on.

10 CFR Part 55 Content: 55.41 7 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-44 Comments /

Reference:

SD000194 Rev: Major: 15 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-44 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-44 Comments /

Reference:

ABN-ELEC-125VDC Rev: Major: 16 Minor: N/A Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-45 Examination Outline Cross-reference: 45 Revision: 0 Date: 8/6/20 Tier: 2 Group: 1 K/A Number: 263000.A3.01 Level of Difficulty: 2 RO Importance Rating: 3.2 K/A

Description:

Ability to monitor automatic operations of the D.C. ELECTRICAL DISTRIBUTION including: Meters, dials, recorders, alarms, and indicating lights CGS is in Mode 1.

A main generator trip occurs coincident with a complete loss of offsite power (LOOP).

With no operator action, how would DC voltage on bus S1-1 and S1-7 compare to the initial values 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after the event?

BUS S1-1 VOLTS DC will be (1) and BUS S1-7 VOLTS DC will be (2) .

A. (1) approximately the same (2) approximately the same B. (1) approximately the same (2) lower C. (1) lower (2) approximately the same D. (1) lower (2) lower Answer: B K/A Match:

Requires knowledge of the change in DC distribution parameters on a LOOP.

Page 1 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-45 Explanation:

A. Incorrect. Plausible since Bus S1-1 voltage will be the same as voltage prior to the event.

However, S1-7 voltage will be lower. See explanation for B.

B. Correct. Battery Charger E-C1-1 normally powers Bus S1-1. It is powered from SM-7 via MC-7A.

On a LOOP with a loss of the Main Generator, Bus SM-7 experiences an Under Voltage (UV) condition. Some loads are shed from the bus. However, MC-7A will not load shed. Approximately 7 seconds after the UV, DG2 will pick up SM-7 and EC1-1 will power Bus S1-1. Bus voltage might lower initial after the event due to the current limiting setting of the battery charger while battery B1-1 is recharging, but one hour after the event, the battery will be recharged and bus voltage will be restored to normal.

Bus S1-7 is powered from Battery Charger E-C1-7. The battery charger is powered from SH-6 via MC-6B. On a LOOP with a loss of the Main Generator, SH-6 will be deenergized. With no operator action, charger E-C1-7 will remain deenergized and bus S1-7 will be powered from the battery. Battery (and bus) voltage will lower as the battery dischargers to supply bus S1-7 loads.

An hour after the event, battery B1-7 will still power the bus and bus voltage will be lower.

C. Incorrect. Plausible if it is believed that MC-7A will be load shed and MC-6B will remain powered during the event described. However, the opposite will be true. See explanation B.

D. Incorrect. Plausible since Bus S1-7 will be lower. However, Bus S1-1 voltage will be approximately the same as before the event. See explanation B.

Technical Reference(s)

SD000188, DC Distribution System Description Attached w/ Revision #

SD000182, AC Distribution System Description See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11838 Describe the effect on the following due to a loss of supporting AC power:

a. Battery indications, discharge rate and capacity

b. DC loads Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Page 2 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-45 Question Cognitive Level:

Justification for Cognitive Level Examinee must synthesize the information given in the stem with an understanding of the results on the AC Distribution system for a LOOP with a loss of Main Generator along with a knowledge of the power supplies to battery chargers.

10 CFR Part 55 Content: 55.41 7 Page 3 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-45 Comments /

Reference:

SD000188 Rev: Major: 10 Minor: 3 Page 4 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-45 Page 5 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-45 Comments /

Reference:

SD000182 Rev: Major: 21 Minor: 3 Page 6 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-45 Page 7 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-46 Examination Outline Cross-reference: 46 Revision: 0 Date: 8/6/20 Tier: 2 Group: 1 K/A Number: 264000.K3.02 Level of Difficulty: 2 RO Importance Rating: 3.9 K/A

Description:

Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: electrical distribution CGS is in Mode 1.

DG-2 has completed a run for maintenance.

DG-2 is restored to a normal standby lineup in accordance with SOP-DG2-STBY, Emergency Diesel Generator (Div 2) Standby Lineup, except the Engine Speed Selector switch, which is left in IDLE.

A LOCA coincident with a Loss of Offsite Power (LOOP) occurs.

Drywell pressure is 1.75 psig, up slow.

How does DG2 respond to this event?

DG2 will start and ramp to (1) speed. Breaker DG2-8 (2) .

A. (1) IDLE (2) will automatically close B. (1) IDLE (2) must be manually closed C. (1) RATED (2) will automatically close D. (1) RATED (2) must be manually closed Answer: C K/A Match:

Requires knowledge of effect on electrical distribution system if DG2 is not in proper lineup.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-46 Explanation:

A. Incorrect. Plausible since DG2 will come up to IDLE speed on a high drywell pressure condition without an UV condition on SM-8. Plausibility is enhanced since DG2-8 will close for the conditions given. However, on a UV on SM-8, DG2 will ramp to RATED speed regardless of Engine Speed Selector switch position.

B. Incorrect. Plausible since DG2 will come up to IDLE speed on a high drywell pressure condition without an UV condition on SM-8. Plausibility is enhanced since DG2-8 will not automatically close if the DG does not come up to RATED speed. However, on a UV on SM-8, DG2 will ramp to RATED speed regardless of Engine Speed Selector switch position, and DG2-8 will automatically close.

C. Correct. On A Loss of Offsite Power (LOOP), SM-8 will see a UV condition. DG2 will start and come up to RATED speed and voltage regardless of Engine Speed Selector switch position. D2-8 will automatically close approximately 7 seconds after the UV condition occurs.

D. Incorrect. Plausible since DG2 will come up to RATED speed with a UV condition on SM-8 regardless of Engine Speed Selector switch position. However, DG2-8 will automatically close 7 seconds after the UV condition occurs.

Technical Reference(s)

SD000200, Emergency Diesel Generator System Description SOP-DG2-STBY, Emergency Diesel Generator (Div 2) Standby Attached w/ Revision #

Lineup See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5321 State the function of the IDLE and RATED positions of the Idling Speed Control switch. Describe the effect of FA or UV signals when in IDLE.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-46 Justification for Cognitive Level Requires knowledge of the effect of Engine Speed Selector switch.

10 CFR Part 55 Content: 55.41 7 Comments /

Reference:

SD000200 Rev: Major: 12 Minor: 5 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-46 Comments /

Reference:

SOP-DG2-STBY Rev: Major: 023 Minor: N/A Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-47 Examination Outline Cross-reference: 47 Revision: 0 Date: 5/13/20 Tier: 2 Group: 1 K/A Number: 300000.K2.01 Level of Difficulty: 2 RO Importance Rating: 2.8 K/A

Description:

Knowledge of electrical power supplies to the following: Instrument air compressor What is the power supply to air compressor CAS-C-1C?

A. MC-2P B. MC-6C C. MC-7A D. MC-8A Answer: A K/A Match:

Requires knowledge of air compressor power supplies.

Explanation:

A. Correct. As listed in SD000205, the power supply to CAS-C-1C is MC-2P.

B. Incorrect. Plausible since MC-6C is the power supply to Service Air Compressor SA-C-1.

C. Incorrect. Plausible since MC-7A is the power supply to air compressor CAS-C-1A.

D. Incorrect. Plausible since MC-8A is the power supply to air compressor CAS-C-1B.

Technical Reference(s)

SD000205, Control and Service Air Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5876 - Identify which Control Air system compressors and Cooling Jacket Water pumps are powered from "Critical" buses.

Page 1 of 2

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-47 Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of air compressor power supplies.

10 CFR Part 55 Content: 55.41 4 Comments /

Reference:

SD000205 Rev: Major: 12 Minor: 1 Page 2 of 2

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-48 Examination Outline Cross-reference: 48 Revision: 0 Date: 5/18/20 Tier: 2 Group: 1 K/A Number: 400000.2.1.27 Level of Difficulty: 2 RO Importance Rating: 3.9 K/A

Description:

Component cooling water: Knowledge of system purpose and/or function.

Which of the following loads are cooled by Reactor Closed Cooling (RCC)?

(1) Condensate Booster Pump bearings (2) Control Rod Drive Pump bearings and oil cooler (3) Reactor Feedwater Pump turbine bearings (4) Offgas glycol refrigeration machines (5) Reactor Recirculation Pump seals (6) Reactor Water Cleanup Regenerative Heat Exchanger A. (1) (2) (3)

B. (1) (3) (6)

C. (2) (4) (5)

D. (4) (5) (6)

Answer: C K/A Match:

Requires knowledge of the purpose of RCC (ie - loads cooled by RCC).

Explanation:

A. Incorrect. Plausible since (2) is cooled by RCC. However, (1) and (3) are cooled by Plant Service Water (TSW).

B. Incorrect. Plausible since (6) is cooled by RCC. However, However, (1) and (3) are cooled by Plant Service Water (TSW).

C. Correct. In accordance with SD000196, (2), (4), and (5) are cooled by RCC.

D. Incorrect. Plausible since (4) and (5) are cooled by RCC, and the RWCU Non-Regenerative Heat Exchanger is cooled by RCC. However, the RWCU Regenerative Heat Exchanger (6) is cooled by RPV return flow.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-48 Technical Reference(s)

SD000196, RCC System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5704 - Identify the cooling loads supplied by the Reactor Building Closed Cooling Water system in the Reactor Building, the Primary Containment and the Radwaste Building.

Question Source: #: LO00788

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of loads supplied by RCC.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-48 Comments /

Reference:

SD000196 Rev: Major: 14 Minor: 2 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-49 Examination Outline Cross-reference: 49 Revision: 0 Date: 06/30/20 Tier: 1 Group: 1 K/A Number: 209002 K1.03 Level of Difficulty: 2 RO Importance Rating: 3.0 K/A

Description:

Knowledge of the physical connections and/or cause and effect relationships between High Pressure Core Spray and the following: Water Leg (jockey)pump With Columbia operating at rated power, the following occurs:

Which of the following describes the impact to the HPCS system?

A. Insufficient NPSH exists for HPCS-P-1 which could result in pump cavitation.

B. Non-condensable gases will not be removed from the HPCS system piping.

C. HPCS-P-1 will not automatically start upon a valid initiation signal.

D. System piping may be damaged due to water hammer if HPCS-P-1 starts.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-49 Answer: D K/A Match:

Requires an understanding of applicable annunciators and the cause and effect relationship between HPCS-P-1 and the water leg pump.

Explanation:

A. Incorrect Plausible. The HPCS water leg pump does provide some value of NPSH to HPCS-P-1.

However, the amount that is provided is not required for proper operation of HPCS-P-1 and operating HPCS-P-1 with the water leg pump unavailable will not result in cavitation of the pump.

HPCS-P-1 is on the reactor building 422 elevation and due to the elevation of the CSTs / Wetwell HPCS-P-1 will have adequate NPSH for operations regardless of the status of HPCS-P-3 B. Incorrect. Plausible. While the keep fill pump maintains the system full such that there is no air or gases in the system upon start, the bases of HPCS-P-3 is not to remove air and non-condensable gases from the system.

C. Incorrect. Plausible.HPCS-P-1 could cause system damage due to water hammer if it auto starts after the keep fill pump has been out of service f, however there is no interlock preventing the automatic start of HPCS-P-1 as a result of the loss of the keep fill pump.

D. Correct. The water leg pump, also called a keep fill pump, is a low capacity, low head pump designed to ensure HPCS discharge piping is maintained full of water. This is designed to minimize injection time and to avoid piping damage from water hammer.

Technical Reference(s)

PPM 4.601.A1 Attached w/ Revision #

SD000174 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 209002 Describe the HPCS system design features or interlocks that provide for the following: a. prevention of water hammer.

Question Source: #: Bank #

(See Below / Modified to include

LO02216 diagnosis)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-49 Justification for Cognitive Level Requires diagnosis / synthesis of HPCS indications and understanding of potential system response based on those indications.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

PPM 4.601.A1 Rev: Major: 32 Minor: N/A Comments /

Reference:

SD000174 Rev: Major: 16 Minor: 2 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-49 Comments /

Reference:

Original Question LO02216 Rev: Major: N/A Minor: N/A Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-50 Examination Outline Cross-reference: 50 Revision: 0 Date: 07/08/20 Tier: 2 Group: 1 K/A Number: 215003 2.1.28 Level of Difficulty: 2 RO Importance Rating: 4.1 K/A

Description:

Knowledge of the purpose and function of major systems components and controls (IRM)

Which one of the following will cause the white INOP light on the IRM drawers (H13-P606 and H13-P633) to illuminate?

The A. associated IRM channel is in BYPASS on the IRM Bypass Joystick.

B. IRM Detector associated with the drawer is not fully inserted in the core.

C. associated IRM High Voltage Power Supply (HVPS) voltage output is low.

D. associated IRM channel is indicating LT 5/125 of scale for the range selected.

Answer: C K/A Match:

Requires an understanding of The IRM system major components and indications.

Explanation:

A. Incorrect. Plausible since an IRM channel that is bypassed does not provide input to RPS or RMCS . However trip and indicating lights on the individual IRM instrument drawers are not affected by the position of the Bypass Joystick.

B. Incorrect. Plausible since a Rod Block is inserted if an IRM detector is not fully inserted when the Reactor Mode Switch is not in RUN. However, detector position does not provide an input to the INOP light.

C. Correct. The INOP light on the individual IRM drawers is illuminated if one of the following occurs: High Voltage Power Supply output voltage is low / any IRM drawer module is unplugged / the Node/Test switch is not in OPERATE / (-) 24 vdc voltage is lost D. Incorrect. Plausible since a reading of LT 5/125 scale will illuminate the Downscale Trip Light. However, it does not provide an input to the INOP light.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-50 Technical Reference(s)

SD000138, IRM Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5449 - Describe the cause and effect relationship between IRM INOP condition and the following.

a. High voltage power supply

b. Voltage preregulator

c. Mode / Test switch

d. IRM Electronics Question Source: #: LO02201

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information given in the stem with an understanding of system response to a loss of HVPS.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-50 Comments /

Reference:

SD000138 Rev: Major: 10 Minor: 2 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-51 Examination Outline Cross-reference: 51 Revision: 0 Date: 07/07/20 Tier: 2 Group: 1 K/A Number: 215004 K4.04 Level of Difficulty: 2 RO Importance Rating: 2.8 K/A

Description:

Knowledge of SOURCE RANGE MONITOR (SRM) SYSTEM design feature(s) and/or interlocks which provide for the following: Changing detector position During reactor start up and power ascension, what actions must be taken to FULLY withdraw the SRM detectors? Why are SRM detectors retracted?

The Detector drive controls must be energized, the SRM select buttons must be lit and the DRIVE OUT pushbutton must be depressed (1) .

SRM detector retraction prolongs the life of the detector by decreasing the level of (2) Flux it is exposed to during reactor operation in the power range.

A. (1) and released (the retraction signal seals in)

(2) Gamma B. (1) continuously until the SRM is fully withdrawn (2) Neutron C. (1) and released (the retraction signal seals in)

(2) Neutron D. (1) continuously until the SRM is fully withdrawn (2) Gamma Answer: B K/A Match:

Requires an understanding of SRM detector controls / interlocks and the bases for SRM retraction at power levels.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-51 Explanation:

A. Incorrect. The DRIVE IN pushbutton is provided with a seal in contact that results in the SRM drive fully inserting the SRM on insert. There is not such seal in for the DRIVE OUT mechanism on the SRMs. Gamma Flux does interact with SRM detector; however, it is not the concern nor part of the bases for retracting the detector during power operations. High neutron flux during power operations will shorten the life of the detector.

B. Correct. The DRIVE OUT mechanism for the SRM detectors does not have a seal in mechanism, the DRIVE OUT push button must be held continuously until the SRM is fully retracted. Detector retraction prolongs the life of the detector by decreasing the level of neutron flux it is exposed to during reactor operation in the power range. It also provides a means of extending the range of detector to provide more overlap between the SRM and IRM Channels.

C. Incorrect. The DRIVE IN pushbutton is provided with a seal in contact that results in the SRM drive fully inserting the SRM on insert. There is not such seal in for the DRIVE OUT mechanism on the SRMs. The second part of this distractor is true, Detector retraction prolongs the life of the detector by decreasing the level of neutron flux it is exposed to during reactor operation in the power range.

D. Incorrect. The first part of the distractor is true, The DRIVE OUT mechanism for the SRM detectors does not have a seal in mechanism, the DRIVE OUT push button must be held continuously until the SRM is fully retracted. The second part of this distractor is false, Gamma Flux does interact with SRM detector; however, it is not the concern nor part of the bases for retracting the detector during power operations. High Neutron flux during power operations will shorten the life of the detector.

Technical Reference(s)

Need a technical reference Attached w/ Revision #

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11997 - Describe the function, purpose and design features of the following major Source Range Monitoring system components. B. Detector insert and retract mechanism.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-51 Justification for Cognitive Level Requires the comprehension of operational for the SRM detector drive controls and the bases of this operational knowledge.

10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

SD000132 Rev: Major: 12 Minor: 1 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-52 Examination Outline Cross-reference: 1 Revision: 0 Date: 07/02/20 Tier: 2 Group: 1 K/A Number: 223002 K6.07 Level of Difficulty: 2 RO Importance Rating: 3.2 K/A

Description:

Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF: Essential AC Power A loss of RPS-A has occurred. Which set of P601 indications is expected for this condition?

A.

B.

C.

D.

Page 1 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-52 Answer: A K/A Match:

Requires an understanding of essential power supplies to the NS4 system and the resulting plant indications if an AC power source is lost including detailed knowledge of inboard and outboard isolations.

Explanation:

A. Correct. A loss of RPS A results in a P601 Outboard Isolation and MSIVs would remain open. As depicted above the Inboard isolation valves FDR-V-3 and RWCU-V-1 Remain open, While the outboard Isolation valves FDR-V-4 and RWCU-V-4 Remain open. Also depicted above, the MSIVs remain open.

B. Incorrect. Plausible. This is a indication of a loss of RPS B which would resulted in both and inboard and outboard isolation with the MSIVs remaining open.

C. Incorrect. Plausible. A loss of PP-7AA (IN-3) would result in the closure of NS4 group 3,4 outboard Isolation valves (as depicted) This effects only FDR-V-4, RWCU-V-4 would still indicate open as it is a group 7 Isolation. MSIVs would remain open.

D. Incorrect. A loss of PP-8AA (IN-2) would result in a closure of NS4 inboard isolation valves. This would only affect FDR-V-3. RWCU-V-4 and FDR-V-4 would remain open. Further, a loss of IN-2 results in isolation of the inboard MSIVs due to how the solenoids receive power. This ultimately results in both inboard and outboard MSIVs going closed due to the pressure sensing point for a group 1 Isolation (831# in run.).

Technical Reference(s)

SD000173 Attached w/ Revision #

ABN-RPS See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5604 - Describe the actions that would occur due to a loss of one or both RPS power supplies to the NS4 logic.

Question Source: #: Bank #

LO00749 (Note changes or attach parent)

Question History: Last NRC Exam: N/A Page 2 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-52 Question Cognitive Level:

Requires diagnosis of actual plant indications and intimate knowledge of NS4 groups including the effects of a loss of RPS and/or UPS to both inboard and outboard NS4 Isolations and MSIVs.

10 CFR Part 55 Content: 55.41 7 Comments /

Reference:

SD000173 Rev: Major: 14 Minor: 5 Page 3 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-52 Comments /

Reference:

ABN-RPS Rev: Major: 15 Minor: N/A Page 4 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-52 Comments /

Reference:

ABN-ELEC-INV Rev: Major: 17 Minor: 1 Page 5 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-52 Page 6 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-52 Comments /

Reference:

LO00740 Rev: Major: N/A Minor: N/A Page 7 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-53 Examination Outline Cross-reference: 53 Revision: 0 Date: 07/01/20 Tier: 2 Group: 1 K/A Number: 262002 K3.14 Level of Difficulty: 3 RO Importance Rating: 2.8 K/A

Description:

Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following: Rx Power A transient occurs that results in the following indication:

What is the expected plant response and mitigation strategy?

A. The plant has entered the OPRM enabled region. Insert Control Rods to reduce Rod Line to less than 100%.

Page 1 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-53 B. The plant has scrammed. Enter PPM 5.1.1 RPV Control. Initial Level control will be with HPCS and RCIC, Pressure control will be with SRVs.

C. The plant remains at 100% power. Enter ABN-RCC and manually scram the reactor on a complete loss of RCC flow.

D. The plant has scrammed. Commence venting hydrogen from the Main Generator.

Manually trip PCB-4885 and PCB-4888 if Main Turbine Vibration increases by GE 2 mils.

Answer: B K/A Match:

Requires diagnosis of plant indication for a loss of UPS power, the resultant plant status and correct initial mitigation strategy outlines in the applicable ABN.

Page 2 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-53 Explanation:

A. Incorrect. Plausible if the operator believes IN-1 Powers Instrument Bus B. The plant would enter the OPRM Enabled region for a loss of IN-1 (which power E-PP-US and NOT E-PP-8AA).

Mitigation strategy for a normal entry into the OPRM enabled region due to a trip pf a RFW pump and RRC runback would be to reduce rodline to LT 100%, however this is also the incorrect mitigation strategy for a loss of IN-1 per ABN-ELEC-INV which requires a scram of the reactor if power could not be restored to IN-1. Both parts of this distractor are incorrect.

B. Correct. Per diagnosis of the provided indications, power has been lost to E-PP-8AA (Loss of IN-2). Per ABN-ELEC-INV:

MSIV closure causes a RPS MSIV Closure scram, when the plant scrams the Crew will have EOP entry into PPM 5.1.1 RPV Control on high reactor pressure. Due to the MSIV closure, initial level control with reactor Feedwater and Condensate will not be available due to no steam to Reactor Feedwater pumps and reactor pressure still greater than the shutoff head of the Condensate Booster pumps. Initial level control will be with RCIC and HPCS. Pressure control will be with the Safety Relief valves. Main Steam Isolation Valve Closure results in bypass valves not being available for pressure control.

C. Incorrect. Plausible is the candidate believes that PP-8AA is associated with IN-3 or does not understand the plant response to a loss of IN-2. A loss of IN-2 or a Loss of IN-3 will result in a complete loss of RCC flow due to the associated Relay Cabinet Outboard (loss of IN-3) and Inboard (Loss of IN-2) Isolations. ABN-ELEC-INV directs the operator to REFER to ABN-RCC for a loss of IN-2 and a loss of IN-3. If IN-3 was lost, the correct mitigation strategy would be to manually scram per ABN-RCC. For a loss of IN-2 the plant has already scrammed on closure of the inboard MSIVs.

D. Incorrect. Plausible if the operator misdiagnoses the indications provided as a loss of DP-S1-7.

The indication for S1-7 voltage is directly above the indication for Instrument Bus B. The mitigation strategy provided in the distractor IS CORRECT for a loss of DP-S1-7. However, DP-S1-7 voltage is normal and has not been lost.

Page 3 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-53 Technical Reference(s)

ABN-ELEC-INV Attached w/ Revision #

ABN-ELEC-125VDC See Comments / Reference ABN-CORE Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5023 - Predict the effect that a loss or malfunction of the following will have on the Reactor Recirculation System: c. Dual RRC Pump trip Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires the operator to diagnose a current plant indication and use that information to determine the resultant plant condition (Power level). It also requires specific ABN mitigation strategy for the situation.

10 CFR Part 55 Content: 55.41 10 Page 4 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-53 Comments /

Reference:

ABN-ELEC-INV Rev: Major: 17 Minor: 1 Plausible Distractor A Plausible Distractor C PER ABN-RCC Major Rev 6 Minor Rev 4 Page 5 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-53 Correct Answer B Comments /

Reference:

ABN-CORE Rev: Major: 17 Minor: 2 Plausible Distractor A Page 6 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-53 Comments /

Reference:

ABN-ELEC-125VDC Rev: Major: 16 Minor: N/A Plausible Distractor D Page 7 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-54 Examination Outline Cross-reference: 54 Revision: 0 Date: 10/28/20 Tier: 2 Group: 2 K/A Number: 201002.A2.01 Level of Difficulty: 4 RO Importance Rating: 2.7 K/A

Description:

Ability to (a) predict the impacts of the following on the REACTOR MANUAL CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Rod movement sequence timer malfunctions CGS is in Mode 1.

Operators are performing OSP-CRD-M701, Control Rod Exercise.

After completing a single notch insertion of a control rod, a fault in the Rod Drive Control System (RDCS) causes the Settle Timer to remain deenergized and the following light to remain lit:

There are no active Rod Insertion Blocks.

How does this affect the RDCS?

The selected control rod (1) insert when the INSERT button is depressed and (2) insert when the CONTINUOUS INSERT button is depressed.

A. (1) will (2) will B. (1) will (2) will not C. (1) will not (2) will D. (1) will not (2) will not Answer: C Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-54 K/A Match:

Requires knowledge of the effects of a failure of the rod movement sequence timer.

Explanation:

A. Incorrect. Plausible since the selected control rod will move when the CONTINUOUS INSERT button is depressed. However, the rod will not move when the INSERT button is depressed.

B. Incorrect. Plausible if it is believed that the Activity Control Timer is used during a continuous insertion. However, depressing the CONTINUOUS INSERT button bypasses the timer to allow rod insertion when the timer malfunctions.

C. Correct. When the timer fails to energize, a control rod insertion block is inserted which prevent the control rod from inserting using the INSERT button. Depressing the CONTINUOUS INSERT button bypasses the timer, allowing control rod insertion when the timer fails.

D. Incorrect. Plausible if it is believed that the Activity Control Timer is active when either button is depressed. However, the function of the CONTINUOUS INSERT button is to bypass the timer allowing rod insertion with a timer failure.

Technical Reference(s)

SD000148, Reactor Manual Control System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5792 State the functions and interrelationships of these P603 controls: b.

Continuous Insert pushbutton.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of effects of Activity Control Timer malfunction and methods available to insert control rods.

Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-54 10 CFR Part 55 Content: 55.41 7 Comments /

Reference:

SD000148 Rev: Major: 16 Minor: 0 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-54 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-55 Examination Outline Cross-reference: 55 Revision: 0 Date: 10/28/20 Tier: 2 Group: 2 K/A Number: 201003.K1.04 Level of Difficulty: 2 RO Importance Rating: 2.9 K/A

Description:

Knowledge of the physical connections and/or cause-effect relationships between CONTROL ROD AND DRIVE MECHANISM and the following: Reactor vessel In reference to the following diagram of the lower reactor vessel:

Stub Tube How is the weight of each Control Rod Drive (CRD) Mechanism supported?

CRD Mechanism weight is supported by the A. Top Guide via the Fuel Assembly.

B. Core Plate via the Fuel Support Piece.

C. Lower RPV Head via the CRD Housing and Stub Tube.

D. Reactor Vessel Support Pedestal via the CRD Support Structure.

Answer: C K/A Match:

Requires knowledge of the interface between the Control Rod Drive Mechanism and the Reactor Vessel.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-55 Explanation:

A. Incorrect. Plausible since the Top Guide provides lateral support for the fuel assemblies and the fuel assemblies rest on the CRD Guide Tube. However, it support the weight of the fuel assemblies or CRDMs.

B. Incorrect. Plausible since the Fuel Support piece supports the weight of the fuel assemblies via the CRD Tube and CRD Housing. However, the core plate only supports the weight of 24 peripheral fuel assemblies that do not have CRDMs.

C. Correct. The CRD Housing is welded to the lower RPV head at the stub tube. It transfers the weight of the fuel assembly above and the CRDM below to the RPV Lower Head.

D. Incorrect. Plausible since the CRD Support Structure will prevent CRDM ejection more than 3 inches. However, under normal hot conditions, there is about 1/4 inch clearance between the CRD Support Structure and the CRD Housing.

Technical Reference(s)

SD000125, Reactor Pressure Vessel System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5003 State the function, purpose and design features of the following Reactor Pressure Vessel components: g. Incore housing and guide tubes.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires understanding of core internals.

10 CFR Part 55 Content: 55.41 2 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-55 Comments /

Reference:

SD000125 Rev: Major: 16 Minor: 2 Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-55 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-55 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-56 Examination Outline Cross-reference: 56 Revision: 0 Date: 6/17/20 Tier: 2 Group: 2 K/A Number: 202002.2.4.2 Level of Difficulty: 2 RO Importance Rating: 4.5 K/A

Description:

Recirculation Flow Control - Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

CGS is in Mode 1.

The operations crew is raising power following a reactor startup from a refueling outage.

Reactor power is 40%.

A reactor scram occurs.

RPV level lowers to -20 inches, then begins to slowly rise.

How do the Reactor Recirculation (RRC) pumps respond to this event?

The RRC pumps will automatically A. trip.

B. runback to 15 Hz.

C. runback to 30 Hz.

D. runback to 51 Hz.

Answer: B K/A Match:

Requires knowledge of RRC pump response to plant conditions that are EOP entry conditions: RPV Level below Level 3 (+13 inches).

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-56 Explanation:

A. Incorrect. Plausible since RRC pumps will automatically trip following a scram above 29.5%

power at the End of Cycle (EOC). However, the question stem indicates that a refueling outage has just been completed. Therefore the RRC EOC-RPT trip is not active.

B. Correct. If RPV Level reaches LE Level 3 (+13 inches) RRC pumps will automatically runback to 15 Hz.

C. Incorrect. Plausible since RRC pumps will runback to 30 Hz when one Reactor Feed Pump trips and RPV level is LE Level 4 (+31.5 inches). However, for the conditions given, RRC pumps will runback to 15 Hz.

D. Incorrect. Plausible since a RRC pump that suffers a loss of one ASD channel will runback to 51 Hz. However, for the conditions given, RRC pumps will runback to 15 Hz.

Technical Reference(s)

SD000184, ASD System Description Attached w/ Revision #

SD000178, RRC System Description See Comments / Reference PPM 5.1.1, RPV Control Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 9684 - Given an initial operating condition, describe the response of the RRFC system to a Level 3 (13") RPV level.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing information given in the question stem with a knowledge of RRC pump runbacks and understanding of the conditions necessary to cause an EOC-RPT trip of the RRC pumps.

10 CFR Part 55 Content: 55.41 7 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-56 Comments /

Reference:

SD000184 Rev: Major: 20 Minor: 2 Comments /

Reference:

SD000178 Rev: Major: Maj Minor: Min Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-56 Comments /

Reference:

PPM 5.1.1 Rev: Major: 20 Minor: 2 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-57 Examination Outline Cross-reference: 57 Revision: 0 Date: 6/17/20 Tier: 2 Group: 2 K/A Number: 214000.K6.02 Level of Difficulty: 2 RO Importance Rating: 2.7 K/A

Description:

Knowledge of the effect that a loss or malfunction of the following will have on the ROD POSITION INFORMATION SYSTEM: Position indication probe CGS is in Mode 2.

Control rod withdrawal for a reactor startup has commenced.

The RO is withdrawing rod 34-11 from position 00 to position 48.

As the rod is notched out from position 40 to position 42, the operator notes the following indications:

X X What has caused these indications?

Control rod 34-11 reed switch for position (1) has failed to (2) .

A. (1) 40 (2) open B. (1) 40 (2) close C. (1) 42 (2) open D. (1) 42 (2) close Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-57 Answer: A K/A Match:

Requires knowledge of the effects of a failed position indication probe on rod position indication.

Explanation:

A. Correct. In accordance with SD000148, RMCS, When a reed switch fails to open during rod movement; i.e., two position signals are sent, the four rod display will show XX for the affected rod. In both cases, the DATA FAULT light will illuminate on H13-P603 and on H13-P615.

B. Incorrect. Plausible since a failure of the position 40 reed switch will cause the indications given.

However, the reed switch must fail to open to cause these indications.

C. Incorrect. Plausible since a reed switch failing to open causes the indications given. However, the position 40 reed switch must fail to open when the rod is at position 42.

D. Incorrect. Plausible since position 42 reed switch failing to close will cause the Data Fault light to energize. However, the four-rod display will be blank for rod 34-11, not XX.

Technical Reference(s)

SD000148, RMCS System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7754 - Predict the effect(s) that a RPIS system failure has on:

a. RMCS Question Source: #: LO00194

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of RPIS failure modes.

10 CFR Part 55 Content: 55.41 6 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-57 Comments /

Reference:

SD000148 Rev: Major: 16 Minor: 0 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-58 Examination Outline Cross-reference: 58 Revision: 0 Date: 6/18/20 Tier: 2 Group: 2 K/A Number: 215002.K4.03 Level of Difficulty: 3 RO Importance Rating: 2.9 K/A

Description:

Knowledge of ROD BLOCK MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following: Initiation point (30%)

CGS is in Mode 1.

APRM #1 is bypassed for emergent repairs.

A non-peripheral control rod is selected.

The in service APRMs read as follows:

APRM #2: 25%

APRM #3: 27%

APRM #4: 29%

What is the condition of both Rod Block Monitor (RBM) channels?

RBM A is (1) , RBM B is (2) .

A. (1) bypassed (2) bypassed B. (1) bypassed (2) NOT bypassed C. (1) NOT bypassed (2) bypassed D. (1) NOT bypassed (2) NOT bypassed Answer: C K/A Match:

Requires knowledge of how each RBM channel receives the Low Power Set Point (LPSP).

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-58 Explanation:

A. Incorrect. Plausible since RBM B receives the LPSP signal from APRM #2 which is below the LPSP of 26%. Therefore, RBM B is bypassed. However, with APRM #1 bypassed, RBM A receives its LPSP signal from APRM #3, which is above the LPSP. Therefore RBM A is NOT bypassed.

B. Incorrect. Plausible if it is believed that the first alternate to RBM A is APRM #2 or #4. Plausibility is enhanced if it is believed that APRM #4 is the primary reference to RBM B. However, the first alternate input to RBM A is APRM #3 (RBM A bypassed) and the primary reference to RBM B is APRM #2 (RBM B NOT bypassed).

C. Correct. With APRM #1 bypassed, the first alternate input is APRM #3. Since APRM #3 is >

LPSP, RBM A is not bypassed. The primary reference for RBM B is APRM #2. Since APRM #2 is < LPSP, RBM B is bypassed.

D. Incorrect. Plausible since RBM A is NOT bypassed. However, RBM B is bypassed.

Technical Reference(s)

SD001819, PRNM System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5085 - Identify which APRM channels are utilized by the rod block monitor channels.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information from the question stem with an understanding of the inputs provided to RBM channels from APRMs along with a knowledge of the Low Power Set Point (LPSP) and how the RBMs are affected when an input is above or below LPSP.

10 CFR Part 55 Content: 55.41 7 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-58 Comments /

Reference:

SD001819 Rev: Major: 4 Minor: 0 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-59 Examination Outline Cross-reference: 59 Revision: 0 Date: 7/29/20 Tier: 2 Group: 2 K/A Number: 233000.A1.03 Level of Difficulty: 2 RO Importance Rating: 3.1 K/A

Description:

Ability to predict and/or monitor changes in parameters associated with operating the FUEL POOL COOLING AND CLEAN-UP controls including: Pool temperature CGS is in Mode 1.

The crew has placed Fuel Pool Heat Exchanger, FPC-HX-1A, out of service for corrective maintenance.

What is the maximum, procedurally allowed, fuel pool temperature under these conditions?

A. 125°F B. 138°F C. 150°F D. 175°F Answer: D K/A Match:

Requires knowledge of the maximum fuel pool temperature allowed when operating controls to place a heat exchanger out of service.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-59 Explanation:

A. Incorrect. Plausible since this is the maximum allowed temperature under normal conditions with both supply water pumps and heat exchangers available. However, fuel pool temperature is allowed to rise to a maximum of 175°F when one supply pump or heat exchanger is unavailable.

B. Incorrect. Plausible since this is the temperature above which SW-V-75AA and SW-V-75BB (SW to FPC Manual Isolation Valves) are required to be open. However, fuel pool temperature is allowed to rise to a maximum of 175°F when one supply pump or heat exchanger is unavailable.

C. Incorrect. Plausible since this is the maximum allowed temperature while draining the Fuel Pool Dryer Separator Pit. However, fuel pool temperature is allowed to rise to a maximum of 175°F when one supply pump or heat exchanger is unavailable.

D. Correct. In accordance with SOP-FPC-OPS, section 4.2, and the FSAR (page 9.1-26), fuel pool temperature is allowed to rise to a maximum of 175°F with one supply pump or heat exchanger unavailable.

Technical Reference(s)

SOP-FPC-OPS, Fuel Pool Cooling and Cleanup Operations Attached w/ Revision #

Final Safety Analysis Report (FSAR)

See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5364 - State the Fuel Pool Cooling and Cleanup System limitation concerning fuel pool temperature.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of fuel pool temperature limits.

10 CFR Part 55 Content: 55.41 10 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-59 Comments /

Reference:

SOP-FPC-OPS Rev: Major: 007 Minor: 008 Comments /

Reference:

FSAR Rev: Major: 007 Minor: 008 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-59 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-60 Examination Outline Cross-reference: 60 Revision: 0 Date: 7/30/20 Tier: 2 Group: 2 K/A Number: 234000.K5.03 Level of Difficulty: 3 RO Importance Rating: 2.9 K/A

Description:

Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLING EQUIPMENT: Water as a shield against radiation CGS is in Mode 5.

Fuel Handling is in progress. The following lights are lit on the Refueling Bridge Controller consoles:

How is the operation of the Refueling Bridge Crane affected?

A. The Main Hoist cannot be raised or lowered.

B. The Main Hoist cannot be lowered, but may be raised.

C. The Main Hoist cannot be raised, but may be lowered.

D. All bridge, trolley and Hoist motion is prevented.

Answer: C K/A Match:

Requires knowledge of Refueling Bridge interlocks to ensure sufficient water over the fuel to prevent high radiation.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-60 Explanation:

A. Incorrect. Plausible since the Main Hoist cannot be raised. However, the Main Hoist may be lowered.

B. Incorrect. Plausible since other interlocks prevent the Main Hoist from being lowered (Slack Cable, Grapple Full Down). However, the Grapple Normal Up interlock prevents the Main Hoist from being raised, but allows the hoist to be lowered.

C. Correct. The Normal Grapple Up interlock stops hoist upward travel to ensure that the top of irradiated fuel is not raised to within 76 of the waters surface.

D. Incorrect. Plausible since the Fault Lockout interlock prevents all bridge, trolley, and hoist motion.

However, the Normal Grapple Up interlock only prevents hoist upward motion.

Technical Reference(s)

SD000207, Fuel Handling System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5358 - Explain the following Refueling Bridge indication: a. GRAPPLE NORMAL UP.

Question Source: #: LO02895

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing information in the stem with an understanding of the indications that the NORMAL GRAPPLE UP interlock is active and a knowledge of the effects of this interlock on the operation of the Refueling Bridge Crane.

10 CFR Part 55 Content: 55.41 13 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-60 Comments /

Reference:

SD000207 Rev: Major: 14 Minor: 0 Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-60 Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-60 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-61 Examination Outline Cross-reference: 61 Revision: 0 Date: 7/30/20 Tier: 2 Group: 2 K/A Number: 239001.K3.15 Level of Difficulty: 2 RO Importance Rating: 3.5 K/A

Description:

Knowledge of the effect that a loss or malfunction of the MAIN AND REHEAT STEAM SYSTEM will have on following: Reactor water level control CGS is in Mode 1.

Reactor power is 100% and steady.

A fault causes the total steam flow input to the Feed Water Level Control (FWLC) system to slightly lower.

FWLC remains in 3 Element Control.

How is the plant affected by this fault?

Initially, Reactor Feed Pump turbine speed will (1) . When compared to conditions prior to the fault, RPV level will stabilize at (2) .

A. (1) remain the same (2) a lower level B. (1) remain the same (2) the same level C. (1) lower (2) a lower level D. (1) lower (2) the same level Answer: C K/A Match:

Requires knowledge of the effects of a Main Steam system fault on RPV level control.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-61 Explanation:

A. Incorrect. Plausible since RFP turbine will stabilize at the same speed as before the fault and the RPV will stabilize at a lower level. However, initially, the steam flow/feed flow mismatch will cause turbine speed to lower.

B. Incorrect. Plausible since RFP turbine will stabilize at the same speed as before the fault.

However, the RPV will stabilize at a lower level.

C. Correct. The FWLC system is level dominant. However, steam flow/feed flow mismatch provides faster response to a change in reactor power. The question stem states that the FWLC system remains in 3 element control. Therefore, steam flow/feed flow provides an input. When the indicated steam flow lowered, the RPV level input to FWLC is lowered. This causes RFP turbine speed to initially lower to reduce actual RPV level to the setpoint. Once RPV level input reaches the setpoint, RFP turbine speed is raised to maintain RPV level at the setpoint. Since actual steam flow did not change, feed flow must be maintained at the initial value prior to the fault, which means that RFP turbine speed will stabilize at the same value as before the fault, with actual RPV level lower than initial level to provide an error signal that is equal to the steam flow/feed flow error caused by the fault.

D. Incorrect. Plausible since RFP turbine speed will initially lower. However, RPV level must stabilize at a lower level to provide an error signal to RWLC that is equal to the steam flow/feed flow error caused by the fault.

Technical Reference(s)

SD000157, Feedwater Level Control System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5395 - Describe the response of the FWLC system during steady-state operation and during a change in reactor power in Single Element and Three Element.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-61 Justification for Cognitive Level Requires knowledge of Feed Turbine response to a steam flow fault along with FLWC response to maintain RPV level.

10 CFR Part 55 Content: 55.41 7 Comments /

Reference:

SD000157 Rev: Major: 19 Minor: 1 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-61 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-62 Examination Outline Cross-reference: 62 Revision: 0 Date: 8/4/20 Tier: 2 Group: 2 K/A Number: 245000.A3.10 Level of Difficulty: 2 RO Importance Rating: 2.5 K/A

Description:

Ability to monitor automatic operations of the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS including: Generator output voltage/reactive load CGS is in Mode 1.

A reactor plant startup is in progress.

Reactor power is 15%.

The operations crew is testing the Main Generator Voltage Adjuster prior to synchronizing the Main Generator to the grid.

The Voltage Regulator control switch is in ON.

Main Generator output voltage is 25 kV.

Main Generator Megavars is 0 MV.

What is the effect if the Main Generator Exciter Voltage Adjuster control switch is taken to RAISE?

Main Generator output voltage will (1) and Main Generator Megavars will (2) .

A. (1) remain the same (2) remain the same B. (1) remain the same (2) increase in the OUT direction C. (1) increase (2) remain the same D. (1) increase (2) increase in the OUT direction Answer: C K/A Match:

Requires knowledge of the automatic changes in output voltage and reactive load when the Voltage Adjuster is repositioned.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-62 Explanation:

A. Incorrect. Plausible since Main Generator Megavars will remain the same. However output voltage will increase.

B. Incorrect. Plausible since these indications are correct if the Main Generator is synchronized to the grid. However, since the Main Generator is not on the grid, output voltage will increase while reactive load (Megavars) will remain at 0.

C. Correct. With the Voltage Regulator ON and the Main Generator not synchronized to the grid, taking the Main Generator Exciter Voltage Adjuster to RAISE will cause Main Generator output voltage to increase. With D. Incorrect. Plausible since output voltage will increase. However, since the Main Generator is not synchronized to the grid, reactive load will remain at 0.

Technical Reference(s)

SD000152 - Main Generator System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7647 - Describe the cause-and-effect relationship between Main Generator MVARs and:

b. Exciter Voltage Adjuster-Raise/Lower Question Source: #: LO03648

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information given in the stem with a knowledge of the operation of the Main Generator voltage regulator and an understanding of how voltage and reactive load will respond when the Main Generator is not synchronized to the grid.

10 CFR Part 55 Content: 55.41 5 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-62 Comments /

Reference:

SD000152 Rev: Major: 13 Minor: 2 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-63 Examination Outline Cross-reference: 63 Revision: 0 Date: 5/6/20 Tier: 2 Group: 2 K/A Number: 256000.K2.01 Level of Difficulty: 2 RO Importance Rating: 2.7 K/A

Description:

Knowledge of electrical power supplies to the following: System pumps CGS is in Mode 1.

A Main Generator trip occurs.

E-SM-2 Startup Power breaker, E-CB-S/2, fails to close.

Which of the following Condensate System pump(s) have power available?

(1) Condensate Pump COND-P-1A (2) Condensate Pump COND-P-1B (3) Condensate Pump COND-P-1C (4) Condensate Booster Pump COND-P-2B A. (1) and (2) ONLY B. (1) and (3) ONLY C. (2) and (4) ONLY D. (3) and (4) ONLY Answer: B K/A Match:

This question requires knowledge of the power supplies to condensate system pumps.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-63 Explanation:

A. Incorrect. Plausible since COND-P-1A is powered from SM-1 and has power available. However, COND-P-1B is powered from SM-2, which is de-energized.

B. Correct. COND-P-1A is powered from SM-1 and COND-P-1C is powered from SM-3. Both of these buses have power.

C. Incorrect. Plausible if it is believed that SM-2 will transfer to the Backup Transformer on a loss of the Startup Transformer. However, for the conditions given in the question stem, SM-2 is de-energized. Therefore, COND-P-1B and COND-P-2B do not have power available.

D. Incorrect. Plausible since COND-P-1C will be available. However, COND-P-2B is powered from E-SM-2. On a loss of E-TR-N1, E-SM-1 through 3 will swap to the Startup Transformer. Since breaker E-CB-S/2 failed to close, E-SM-2 will lose power and COND-P-2B will not be available.

Technical Reference(s)

SD000182, AC Distribution Attached w/ Revision #

SD000134, Condensate See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5058 - Identify the loads on the following buses: a. SM-1, SM-2, SM-3 Question Source: #: NRC 2019 Exam

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: 2019 Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of power supplies to condensate system pumps.

10 CFR Part 55 Content: 55.41 4 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-63 Comments /

Reference:

SD000182 Rev: Major: 21 Minor: Min Comments /

Reference:

SD000134 Rev: Major: 17 Minor: 3 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-64 Examination Outline Cross-reference: 64 Revision: 0 Date: 8/4/20 Tier: 2 Group: 2 K/A Number: 259001.A1.01 Level of Difficulty: 2 RO Importance Rating: 3.1 K/A

Description:

Ability to predict and/or monitor changes in parameters associated with operating the REACTOR FEEDWATER SYSTEM controls including: Feedwater flow/pressure CGS is in Mode 1.

A reactor plant startup is in progress.

RFW-P-1B is being placed in service as the second Reactor Feed Pump in accordance with SOP-RFT-START, Reactor Feedwater Turbine System Start.

Plant conditions:

RFW-P-1A is in AUTO.

RFW-P-1B is in MDEM with Pump Discharge Valve, RFW-W-102B, open.

RPV level is being maintained with the RPV Master Level Controller, RFW-LIC-600, in AUTO.

RFW-P-1A and RFW-P-1B discharge pressures are equal and both RFW pumps are feeding.

CRO1 depresses the UP arrow for RFW-P-1B on RFT-COMP-1.

How are both Reactor Feed Pumps affected?

RFW-P-1B speed will (1) . RFW-P-1A speed will (2) .

A. (1) rise (2) lower B. (1) rise (2) remain the same C. (1) remain the same (2) rise D. (1) remain the same (2) remain the same Answer: A Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-64 K/A Match:

Requires knowledge of the effect of changing RFP settings on speed (i.e. flow) of both pumps.

Explanation:

A. Correct. Pressing the UP arrow on RFT-COMP-1 will cause the speed/flow of RFW-B-1B to rise.

This action does not affect RFW-P-1A directly. The FWLC system will detect the rise in feed flow and signal feed pumps to lower flow by reducing speed. Since RFW-P-1A is in AUTO, its speed/flow will lower. However, since RFW-P-1B is in MDEM, it will not respond to this signal and will remain at a speed/flow higher than the initial value.

B. Incorrect. Plausible since RFW-P-1B speed/flow will increase. However, since RFW-P-1A is in AUTO, it will respond to the FWLC system which sends a signal to slow the feed pumps. See A above.

C. Incorrect. Plausible if it is believed that the UP arrow on RFW-COMP-1 only affects any feed turbine in AUTO, similar to the operation of Master Level Controller controls. However, pressing Pressing the UP arrow on RFT-COMP-1 will cause the speed/flow of RFW-B-1B to rise. See A above.

D. Incorrect. Plausible if it is believed that The UP arrow on RFW-COMP-1 only affects an individual feed turbine if it is in AUTO. However, pressing Pressing the UP arrow on RFT-COMP-1 will cause the speed/flow of RFW-B-1B to rise. See A above.

Technical Reference(s)

SD000157, Feedwater Level Control System Description Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5394 - Describe the function of each of the following controls and how they relate to each other:

a. Startup Level Controller

b. Master Controller

c. Turbine Speed Controllers

d. RFP Minimum Flow Controllers Question Source: #: LO03701

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-64 Justification for Cognitive Level Requires examinee to synthesize information given in the question stem with a knowledge of the operation of the Master Level Controller and an understanding of how adjusting one feed turbine will affect the operation of the other feed turbine in auto.

10 CFR Part 55 Content: 55.41 4 Comments /

Reference:

SD000157 Rev: Major: 19 Minor: 1 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-64 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-65 Examination Outline Cross-reference: 65 Revision: 0 Date: 5/6/20 Tier: 2 Group: 2 K/A Number: 268000.A4.01 Level of Difficulty: 3 RO Importance Rating: 3.4 K/A

Description:

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

CGS is in Mode 1.

A leak in the Reactor Building is filling Sump FDR-SUMP-R1.

Currently, the condition of the sump timer, FDR-TM-605A is as follows:

When will the timer cause an alarm?

The alarm will sound if the sump pump starts before the (1) needle reaches (2) minutes.

A. (1) black (2) 0 B. (1) red Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-65 (2) 0 C. (1) black (2) 20 D. (1) red (2) 30 Answer: B K/A Match:

Requires knowledge of the operation of Reactor Building sump timers.

Explanation:

A. Incorrect. Plausible since the timer moves towards 0 minutes. However, the black needle represents the setpoint and the red needle moves from the setpoint to 0 minutes.

B. Correct. When the timer is reset, the red and black needles are on the same number. The red needle times out by moving towards 0 minutes. The graphic in the question shows that the timer was set at 20 minutes. The timer has been running approximately 10 minutes without the sump pump starting. If the sump pump starts before the red needle times down to 0 minutes, an alarm will sound to signify that the sump s filling faster than expected.

C. Incorrect. Plausible if it is believed that the black needle travels towards 0 minutes when the timer is running and that the red needle represents the alarm setpoint. However, the black needle represents the timer setpoint that is manually inserted. The red needle moves from the setpoint (black needle position) to 0 minutes.

D. Incorrect. Plausible since the red needle moves. however, the red needle moves from the setpoint (black needle position) to 0 minutes. If the sump pump starts prior to the red needle reaching 0 minutes, an alarm will sound.

Technical Reference(s)

SD000167, Leak Detection Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: None Learning Objective: 5466 - Sate the purpose of the leak detection system/s sump pump-out and fill rate timers.

Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-65 Question Source: #: 2019 NRC Exam

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: 2019 Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the operation of the sump timers.

10 CFR Part 55 Content: 55.41 13 Comments /

Reference:

SD000167 Rev: Major: 13 Minor: 2 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-66 Examination Outline Cross-reference: 66 Revision: 0 Date: 6/29/20 Tier: 3 Group: N/A K/A Number: 2.1.2 Level of Difficulty: 2 RO Importance Rating: 4.1 K/A

Description:

Knowledge of operator responsibilities during all modes of plant operation.

CGS is in Mode 1.

The operations crew needs to lower recirculation flow as part of a surveillance.

With CRS/SM permission and under direct supervision of a licensed RO, which of the following individual may operate the RRC pump controls to lower flow?

(1) An unlicensed individual currently in training for a SRO license.

(2) A previously licensed RO that does not currently hold a license.

(3) An operator with an INACTIVE RO license that is standing an under-instruction watch for license activation.

A. (1) ONLY B. (2) ONLY C. (1) and (3) ONLY D. (2) and (3) ONLY Answer: C K/A Match:

Requires knowledge of operators that are allowed to perform reactivity manipulations.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-66 Explanation:

A. Incorrect. Plausible since an individual in license training is authorized to manipulate control room controls under the direct supervision of a licensed operator. However, (3) is also correct.

B. Incorrect. Plausible since a previously licensed individual has performed reactivity manipulations in the past and should possess the requisite knowledge to complete the task correctly. However, non-licensed individuals may not operate control room controls unless in an emergency when directed or if the individual is in a training program leading to a license.

C. Correct. In accordance with PPM 1.3.1, Operating Policies, Programs and Practices, section 4.6.3, individuals in a training program for an operator license may operate control room controls to perform a reactivity manipulation while under the direct supervision of a licensed operator. In addition, inactive licensed SROs and ROs may perform reactivity manipulations while directly supervised by a licensed operator.

D. Incorrect. Plausible (3) is correct. However, an individual not in a training program for a license and without a license (active or inactive) may not perform reactivity manipulations.

Technical Reference(s)

PPM 1.3.1, Operating Policies, Programs and Practices Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 6076 - Identify who can manipulate Control Room controls.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of individuals authorized to perform reactivity manipulations in the control room.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-66 Comments /

Reference:

PPM 1.3.1 Rev: Major: 1281 Minor: N/A Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-67 Examination Outline Cross-reference: 67 Revision: 0 Date: 6/23/20 Tier: 3 Group: N/A K/A Number: 2.1.20 Level of Difficulty: 2 RO Importance Rating: 4.6 K/A

Description:

Ability to interpret and execute procedure steps.

An operator has been directed to lineup a system for startup.

The procedure to be used is designated as Reference Use.

How should the operator placekeep the procedure?

The operator...

A. is not required to placekeep this procedure.

B. should placekeep as often as practical (the end of a section, prior to a break, etc.).

C. must mark each step as complete prior to proceeding to the next step.

D. should print their name and place their initials on the procedure cover sheet to signify that all steps were properly completed.

Answer: B K/A Match:

Requires knowledge of methods to complete Reference Use procedures.

Explanation:

A. Incorrect. Plausible since placekeeping is not required for Information Use procedures. However, placekeeping is required on Reference Use procedures B. Correct. In accordance with SWP-PRO-01, Procedure and Work Instruction Use and Adherence, section 4.5.6: IF using a Reference Use procedure or work instructions, THEN PLACEKEEP steps as often as practical (e.g., after each section is complete, prior to a break, etc.).

C. Incorrect. Plausible since this method is required when performing most continuous use procedures. However, it is not required to be performed on Reference Use procedures after each step.

D. Incorrect. Plausible since some Continuous Use procedures require a printed name and signature on the cover sheet. However, this method is not used on Reference Use procedures.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-67 Technical Reference(s)

SWP-PRO-01, Procedure and Work Instruction Use and Adherence Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 16047 - Upon Completion of this class, the students will be able to demonstrate the ability to perform Licensed Operator tasks while in compliance with procedures to meet management expectations.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information in the question with a knowledge of placekeeping methods along with requirements for Reference Use procedures.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-67 Comments /

Reference:

SWP-PRO-01 Rev: Major: 033 Minor: N/A Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-68 Examination Outline Cross-reference: 68 Revision: 0 Date: 8/11/20 Tier: 3 Group: N/A K/A Number: 2.1.42 Level of Difficulty: 2 RO Importance Rating: 2.5 K/A

Description:

Knowledge of new and spent fuel movement procedures.

CGS is in Mode 5.

Core refueling is in accordance with PPM 6.3.2, Fuel Shuffling and/or Offloading and Reloading.

The crew is preparing to a fuel cell into the core.

Which of the following conditions must be met prior to loading fuel into the core?

(1) No personnel allowed in the drywell.

(2) All control rods fully inserted in fueled cells.

(3) Continuous communications established between the refueling floor and control room.

A. (1) ONLY B. (2) ONLY C. (1) and (3) ONLY D. (2) and (3) ONLY Answer: D K/A Match:

Requires knowledge of requirements to move fuel into the core.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-68 Explanation:

A. Incorrect. Plausible since Drywell entry may be restricted during certain evolutions. However, personnel are allowed to be in the drywell while loading fuel into the core.

B. Incorrect. Plausible since (2) is a requirement to load fuel into the core. However, it is not the only requirement listed.

C. Incorrect. Plausible since (3) is a requirement to load fuel into the core. However, (1) is not a requirement.

D. Correct. In accordance with PPM 6.3.2, Fuel Shuffling and/or Offloading and Reloading, Attachment 8.2, Criteria For Stopping Fuel Loading, both communications with the refueling floor and all control rods fully inserted in fueled cells are required to load fuel in the core.

Technical Reference(s)

PPM 6.3.2, Fuel Shuffling and/or Offloading and Reloading Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8830 - Discuss what actions are required if communication is lost between the bridge and the control room during fuel shuffling.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires Knowledge of requirements for loading fuel.

10 CFR Part 55 Content: 55.41 13 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-68 Comments /

Reference:

PPM 6.3.2 Rev: Major: Maj Minor: Min Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-68 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-69 Examination Outline Cross-reference: 69 Revision: 0 Date: 6/23/20 Tier: 3 Group: N/A K/A Number: 2.2.22 Level of Difficulty: 2 RO Importance Rating: 4.0 K/A

Description:

Knowledge of limiting conditions for operations and safety limits.

Which of the following is the lowest reactor steam dome pressure that results in exceeding Technical Specification Safety Limits?

A. 1310 psig B. 1320 psig C. 1330 psig D. 1340 psig Answer: C K/A Match:

Requires knowledge of Safety Limits.

Explanation:

A. Incorrect. Plausible since this pressure is above the RPS High RPV Pressure Scram setpoint and within 15 psig of the Safety Limit. However, the Safety Limit is 1325 psig, which is above this pressure.

B. Incorrect. Plausible since this pressure is within 5 psig of the Safety Limit. However, the Safety Limit is 1325 psig, which is above this pressure.

C. Correct. 1330 psig is the lowest pressure listed that exceeds the Safety Limit of 1325 psig.

D. Incorrect. Plausible since this pressure exceeds the Safety Limit. However, it is not the lowest pressure listed that exceeds the safety limit.

Technical Reference(s)

Technical Specifications Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-69 Proposed references to be provided during examination: N/A Learning Objective: 10304 - Describe each of the Safety Limits and state what actions are required if a Safety Limit is violated.

Question Source: #: LO02588

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the Safety Limit value for steam dome pressure.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-69 Comments /

Reference:

Technical Specifications Amendment 254 Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Examination Outline Cross-reference: 70 Revision: 0 Date: 8/4/20 Tier: 3 Group: N/A K/A Number: 2.2.39 Level of Difficulty: 3 RO Importance Rating: 3.9 K/A

Description:

Knowledge of less than or equal to one hour Technical Specification action statements for systems CGS is in Mode 1.

A seismic event causes the following:

DG-1 is inoperable.

The Ashe 230 kV line is lost.

The CRS enters LCO 3.8.1, AC Sources - Operating.

What is the earliest action required by LCO 3.8.1?

A. Restore DG-1 to operable.

B. Restore the Ashe 230 kV line to operable.

C. Enter LCO 3.0.3 and prepare for a reactor shutdown.

D. Complete OSP-ELEC-W101, Offsite Station Power Alignment Check.

Answer: D K/A Match:

Requires knowledge of the one hour LCO action for a loss of an offsite circuit.

Page 1 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Explanation:

A. Incorrect. Plausible since the conditions in the question requires entry into LCO 3.8.1, Condition D, which requires restoring the DG OR the Offsite Source within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . However, Conditions A and B are also entered. Both of these conditions require that SR 3.8.1.1 (OSP-ELEC-W101) be completed within one hour.

B. Incorrect. Plausible since the conditions in the question requires entry into LCO 3.8.1, Condition D, which requires restoring the DG OR the Offsite Source within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . However, Conditions A and B are also entered. Both of these conditions require that SR 3.8.1.1 (OSP-ELEC-W101) be completed within one hour.

C. Incorrect. Plausible since LCO 3.8.1, Condition G (Three or more required AC sources inoperable) requires entry into LCO 3.0.3 immediately. Plausibility is further enhanced since LCO 3.8.1, Condition D (which should be entered) requires entry into LCO 3.8.7 if any division has no AC sources. LCO 3.8.7, Condition F requires immediate entry into LCO 3.0.3. However, LCO 3.8.7 is not entered and there are only two AC sources lost. For the conditions given, LCO 3.0.3 would not be entered.

D. Correct. For the conditions given, LCO 3.8.1, Conditions A, B, and D are entered. Both Conditions A and B require completing SR 3.8.1.1 (OSP-ELEC-W101) within one hour.

Technical Reference(s)

Technical Specification LCO 3.8.1, AC Sources - Operating Attached w/ Revision #

OSP-ELEC-W101, Offsite Station Power Alignment Check See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5059 - Referencing Technical Specifications associated with the AC Distribution System and a set of plant conditions, determine as applicable the LCO, the action statement, and the appropriate bases.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 2 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Justification for Cognitive Level Requires synthesizing information in the question stem with a knowledge of specific conditions of Technical Specification Action Statements couple with a knowledge of less than or equal to one hour actions.

10 CFR Part 55 Content: 55.41 10 Page 3 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Comments /

Reference:

TS LCO 3.8.1 Rev: Major: 225 Minor: N/A Page 4 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Page 5 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Page 6 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-70 Comments /

Reference:

OSP-ELEC-W101 Rev: Major: 032 Minor: 004 Page 7 of 7

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-71 Examination Outline Cross-reference: 71 Revision: 0 Date: 6/23/20 Tier: 3 Group: N/A K/A Number: 2.3.13 Level of Difficulty: 3 RO Importance Rating: 3.4 K/A

Description:

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 CGS is in Mode 1.

An operator needs to enter an area to reposition a valve.

The highest dose rate in the area is 525 Rad/Hr at 1 meter.

Emergency conditions DO NOT exist.

Who must authorize this entry?

The entry must be authorized by the A. Shift Manager (SM) ONLY.

B. Radiation Protection Manager (RPM) ONLY.

C. Plant General Manager (PGM) ONLY.

D. Radiation Protection Manager (RPM) and Plant General Manager (PGM).

Answer: D K/A Match:

Requires knowledge of radiological safety procedures for entering Very High Radiation Areas.

Page 1 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-71 Explanation:

A. Incorrect. Plausible since the shift manager may authorize entry into a VHRA during emergency conditions. However, the stem states that emergency conditions do not exist.

B. Incorrect. Plausible since the RPM must authorize the entry. However, this is not the only authorization that is required to enter a VHRA for the conditions given.

C. Incorrect. Plausible since the RPM must authorize the entry. However, this is not the only authorization that is required to enter a VHRA for the conditions given.

D. Correct. In accordance with PPM 11.2.7.3, High Radiation Area, Locked High Radiation Area, and Very High Radiation Area Controls, the RPM and PGM must authorize entry into a VHRA for operational reasons.

Technical Reference(s)

PPM 11.2.7.3, High Radiation Area, Locked High Radiation Area, and Very High Radiation Area Controls Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 11257 - Knowledge of 10CFR: 20 and related facility radiation control requirements.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information in the question with a knowledge of the definition of a VHRA along with an understanding of the requirements to enter a VHRA.

10 CFR Part 55 Content: 55.41 12 Page 2 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-71 Comments /

Reference:

PPM 11.2.7.3 Rev: Major: 042 Minor: 001 Page 3 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-71 Page 4 of 4

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-72 Examination Outline Cross-reference: 72 Revision: 0 Date: 6/23/20 Tier: 3 Group: N/A K/A Number: 2.3.11 Level of Difficulty: 2 RO Importance Rating: 3.8 K/A

Description:

Ability to control radiation releases.

CGS is in Mode 1.

A leak occurs from the primary into secondary containment.

The CRS enters PPM 5.3.1, Secondary Containment Control.

The leak is NOT isolable.

Which of the following combinations of parameters, exceeding Maximum Safe Operating Value (MSOV), will require the crew to perform an emergency depressurization (ED)?

(1) RHR Pump A Room Temperature (2) RHR Pump B Room Level (3) RHR Pump C Room Temperature (4) RHR Pump A Room Radiation Level A. (1) and (2)

B. (1) and (3)

C. (2) and (4)

D. (3) and (4)

Answer: B K/A Match:

Requires knowledge of actions taken to control radiation release.

Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-72 Explanation:

A. Incorrect. Plausible since 2 areas need to reach MSOV to require an ED. However, to ED there needs to be two areas of the same parameter (temperature, level, radiation level) that reach MSOV.

B. Correct. In accordance with step SC-15 of PPM 5.3.1, when any one parameter exceeds its MSOV in two or more areas, an ED is required.

C. Incorrect. Plausible since two parameters have reached MSOV in two separate areas. However, to ED there needs to be two areas of the same parameter (temperature, level, radiation level) that reach MSOV.

D. Incorrect. Plausible since two parameters have exceeded MSOV in the same area. However, to ED there needs to be two areas of the same parameter (temperature, level, radiation level) that reach MSOV.

Technical Reference(s)

PPM 5.3.1, Secondary Containment Control Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8459 - Given a list, identify the statement that describes the two reasons for emergency depressurizing the RPV if one secondary containment parameter is above Maximum Safe Operating Levels in more than one area and a primary system is discharging reactor coolant into secondary containment.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing information given in the stem with a knowledge of the different paths of PPM 5.3.1 along with an understanding of the requirements to perform an emergency depressurization.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-72 Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor: N/A Page 3 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 Examination Outline Cross-reference: 73 Revision: 0 Date: 8/5/20 Tier: 3 Group: N/A K/A Number: 2.4.4 Level of Difficulty: 2 RO Importance Rating: 4.5 K/A

Description:

Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

CGS is in Mode 1.

A slow increase in drywell unidentified leakage is raising drywell pressure.

Drywell pressure is 0.7 psig, up slow.

A loss of IN-1 occurs, resulting in a loss of control rod position indication.

10 seconds later, Drywell pressure reaches 1.0 psig and the CRS directs a manual reactor scram.

Current plant conditions:

Reactor power: 2%, down slow.

RPV level: -15 inches, up slow.

Drywell pressure: 1.1 psig, up slow.

Which EOP(s) is/are required to be entered?

Enter PPM 5.1.1, RPV Control A. ONLY.

B. and PPM 5.2.1, Primary Containment Control.

C. and PPM 5.3.1, Secondary Containment Control.

D. and transition to PPM 5.1.2, RPV Control - ATWS.

Answer: D K/A Match:

Requires knowledge of EOP entry requirements.

Page 1 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 Explanation:

A. Incorrect. Plausible since PPM 5.1.1. should be entered due to low RPV level. However, a loss of the full core display occurs due to the loss of E-US-PP. Without the display, the operators cannot verify that existing control rod pattern is sufficient to ensure the reactor remains shutdown (step RC-2 of PPM 5.1.1). Therefore PPM 5.1.2 should be entered.

B. Incorrect. Plausible since PPM 5.1.1. should be entered due to low RPV level. Additionally, entry into PPM 5.2.1 is plausible based on rising drywell pressure. However, PPM 5.2.1 is not required to be entered until drywell pressure reaches 1.68 psig.

C. Incorrect. Plausible since PPM 5.1.1. should be entered due to low RPV level. Additionally, entry into PPM 5.3.1 is plausible based on the loss of several radiation monitors with the loss of E-US-PP. However, PPM 5.3.1 is entered on radiation monitors in alarm, not with a loss of power.

D. Correct. PPM 5.1.1 is entered due to low RPV water level. Step RC-2 of PPM 5.1.1 states that if it is determined that existing control rod pattern alone does not always assure reactor shutdown, then a transition to PPM 5.1.2 is required. Once in PPM 5.1.2, operators verify that control rods are sufficient to maintain the reactor shutdown by performing SOP-RXSD-DETERMINATION-QC.

Operators may then exit PPM 5.1.2 (see step RC-2) and re-enter PPM 5.1.1.

Technical Reference(s)

ABN-ELEC-INV, 120 VAC Critical Distribution System Failures PPM 5.1.1, RPV Control Attached w/ Revision #

PPM 5.1.2, RPV Control - ATWS See Comments / Reference SOP-RXSD-DETERMINATION-QC, Reactor Shutdown Determination Quick Card Proposed references to be provided during examination: N/A Learning Objective: 6827 - Given a Loss of IN-1, IN-2A/B, IN-3A/B or IN-5, identify those automatic actions that may have occurred.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires understanding of the consequences of a loss of E-US-PP along with knowledge of conditions requiring entry into EOPs.

Page 2 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 10 CFR Part 55 Content: 55.41 10 Comments /

Reference:

ABN-ELEC-INV Rev: Major: 017 Minor: 001 Page 3 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 Comments /

Reference:

PPM 5.1.1 Rev: Major: 22 Minor: N/A Page 4 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 Comments /

Reference:

PPM 5.1.2 Rev: Major: 26 Minor: N/A Page 5 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-73 Comments /

Reference:

SOP-RXSD-DETERMINATION-QC Rev: Major: 001 Minor: N/A Page 6 of 6

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-74 Examination Outline Cross-reference: 74 Revision: 0 Date: 8/5/20 Tier: 3 Group: N/A K/A Number: 2.4.6 Level of Difficulty: 2 RO Importance Rating: 3.7 K/A

Description:

Knowledge of EOP mitigation strategies.

CGS is in Mode 1.

An automatic scram signal is received. Control Rods did not insert.

The CRS enters PPM 5.1.1, RPV Control, and transitions to PPM 5.1.2, RPV Control - ATWS.

Current plant conditions:

Reactor power: 16% down slow.

RPV level: -7 inches, up slow.

MSIVs are closed.

How should the crew proceed in PPM 5.1.2?

The crew should perform the (1) leg of PPM 5.1.2 first to (2) .

A. (1) level (2) ensure that level does not go below Top of Active Fuel (TAF)

B. (1) level (2) reduce the probability of fuel damaging power oscillations C. (1) pressure (2) ensure that pressure remains less than the high pressure scram setpoint D. (1) pressure (2) reduce the probability of equipment damage due to SRV cycling.

Answer: B K/A Match:

Requires knowledge of EOP mitigation strategy during an ATWS.

Page 1 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-74 Explanation:

A. Incorrect. Plausible since the level leg of PPM 5.1.2 is the first leg of the EOP to be performed.

However, the level leg does not necessarily maintain RPV level above TAF (-161 inches).

B. Correct. In accordance with OI-15, EOP and EAL Clarifications, Due to the higher rodline associated with MELLA+, a level leg first strategy during ATWS conditions will reduce the probability of fuel damaging power oscillations. Level should be lowered promptly and communications minimized until the order to stop and prevent is issued.

C. Incorrect. Plausible since one of the purposes of the pressure leg of PPM 5.1.2 is to reduce RPV pressure below the high pressure scram setpoint to avoid SRVs lifting. However, this is not the first procedure leg to be performed. See B above.

D. Incorrect. Plausible since the question stem states that MSIVs are closed. Therefore, it is important to reduce pressure to prevent equipment damage due to SRVs cycling. However, this is not the first procedure leg to be performed. See B above.

Technical Reference(s)

OI-15, EOP and EAL Clarifications Attached w/ Revision #

PPM 5.1.2, RPV Control - ATWS See Comments / Reference Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 13568 - Given a copy of PPM 5.1.2 RPV Control - ATWS and an event requiring entry into the EOPs, execute the strategies of this PPM in accordance with procedure use standards and expectations without error.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to have knowledge of the first procedure leg in PPM 5.1.2 to be performed, and the reason for performing this leg first.

10 CFR Part 55 Content: 55.41 10 Page 2 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-74 Comments /

Reference:

OI-15 Rev: Major: 032 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-74 Comments /

Reference:

PPM 5.1.2 Rev: Major: 26 Minor: N/A Page 4 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-74 Page 5 of 5

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-75 Examination Outline Cross-reference: 75 Revision: 0 Date: 6/23/20 Tier: 3 Group: N/A K/A Number: 2.4.16 Level of Difficulty: 2 RO Importance Rating: 3.5 K/A

Description:

Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

CGS is in Mode 1.

An event occurs that requires control room evacuation.

The CRS enters ABN-CR-EVAC, Control Room Evacuation and Remote Cooldown.

All immediate actions are completed prior to leaving the control room.

Current plant conditions:

One control rod is at position 06. All other control rods are fully inserted.

RPV level is -10 inches, up slow.

Drywell pressure is 1.6 psig, up slow.

Reactor Building differential pressure is 0 inches H2O, and stable.

Which procedure should be used for overall control of the plant?

A. PPM 5.1.1, RPV Control.

B. PPM 3.3.1, Reactor Scram.

C. PPM 5.3.1, Secondary Containment Control.

D. ABN-CR-EVAC, Control Room Evacuation and Remote Cooldown.

Answer: D K/A Match:

Requires knowledge of coordination between EOPs and ABN-CR-EVAC Page 1 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-75 Explanation:

A. Incorrect. Plausible since entry conditions of PPM 5.1.1 are met (RPV level LE +13 inches).

However, a note in ABN-CR-EVAC states that this procedure supercedes EOP procedures.

B. Incorrect. Plausible since this procedure should be entered for a reactor scram and scramming the reactor is an immediate action of ABN-CR-EVAC. However, a note in ABN-CR-EVAC states that this procedure supercedes PPM 3.3.1.

C. Incorrect. Plausible since the entry condition of PPM 5.3.1 have been met (RB d/p LE 0 inches).

However, a note in ABN-CR-EVAC states that this procedure supercedes PPM 3.3.1.

D. Correct. Normally, EOPs have precedence over ABNs. However, a note in ABN-CR-EVAC states that this procedure supercedes EOPs and PPM 3.3.1 Technical Reference(s)

ABN-CR-EVAC Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 6105 - State which procedures have priority/precedence over all other operating procedures when an emergency exists.

Question Source: #: LO01730

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information in the question with a knowledge of EOP entry conditions and a knowledge of notes in ABN-CR-EVAC.

10 CFR Part 55 Content: 55.41 10 Page 2 of 3

ES-401 CGS NRC 2021 RO Written Exam Worksheet Form ES-401-5 Question: RO-75 Comments /

Reference:

ABN-CR-EVAC Rev: Major: 042 Minor: N/A Page 3 of 3

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Examination Outline Cross-reference: 76 Revision: 0 Date: 11/30/20 Tier: 1 Group: 1 K/A Number: 295003.2.4.20 Level of Difficulty: 3 SRO Importance Rating: 4.3 K/A

Description:

Partial or complete loss of AC power: Knowledge of the operational implications of EOP warnings, cautions, and notes.

CGS is in Mode 1.

A Station Blackout (SBO) occurs.

DG-1 and DG-2 will not be available for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

The CRS enters PPM 5.6.1, SBO/ELAP.

Station battery loads have been reduced.

Offsite power will not be available for 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months .

DG-3 and DG-4 are available.

What is the preferred method for restoring AC buses?

The CRS should direct aligning DG-3 to power (1) and aligning DG-4 to power (2) .

A. (1) Division 1 critical switchgear via SM-7 (2) Division 2 critical switchgear via SM-8 B. (1) Division 1 critical switchgear via SM-7 (2) Division 2 critical switchgear via MC-8A C. (1) Division 2 critical switchgear via SM-8 (2) Division 1 critical switchgear via SM-7 D. (1) Division 2 critical switchgear via SM-8 (2) Division 1 critical switchgear via MC-7A Answer: D K/A Match:

Requires knowledge of notes in EOP supplemental procedure PPM 5.6.2.

Page 1 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 SRO Only:

Page 2 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Explanation:

A. Incorrect. Plausible since PPM 5.6.2, SBO and ELAP Attachments, attachments 8.7 and 8.8 allow DG-3 to power SM-7 and DG-4 to power SM-8 if necessary. However, the note prior to step 8.7.3 directs that it is preferred to align DG-3 to power Div. 2 switchgear ( via SM-8) and the note prior to step 8.8.1 directs that it is preferred to align DG-4 to Div. 1 switchgear (via MC-7A).

B. Incorrect. Plausible since PPM 5.6.2, SBO and ELAP Attachments, attachments 8.7 and 8.8 allow DG-3 to power SM-7 and DG-4 to power SM-8 if necessary. However, the note prior to step 8.7.3 directs that it is preferred to align DG-3 to power Div. 2 switchgear ( via SM-8) and the note prior to step 8.8.1 directs that it is preferred to align DG-4 to Div. 1 switchgear (via MC-7A).

C. Incorrect. Plausible since the note prior to step 8.7.3 directs that it is preferred to align DG-3 to power Div. 2 switchgear (via SM-8) and the note prior to step 8.8.1 directs that it is preferred to align DG-4 to Div. 1 switchgear. However, DG-4 is not connected to SM-7 directly, it must be connected via MC-7A.

D. Correct. Although both DGs are capable of powering either Div. 1 or Div. 2 critical switchgear, the note prior to step 8.7.3 directs that it is preferred to align DG-3 to power Div. 2 switchgear ( via SM-8) and the note prior to step 8.8.1 directs that it is preferred to align DG-4 to Div. 1 switchgear (via MC-7A).

Technical Reference(s)

PPM 5.6.1, SBO/ELAP Attached w/ Revision #

PPM 5.6.2, SBO and ELAP Attachments See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 13570 - Given a copy of PPM 5.6.1, Station Blackout and an event requiring entry into the EOPs, execute the strategies of thisPPM in accordance with procedure use standards and expectations without error.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Page 3 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Justification for Cognitive Level Requires examinee to synthesize information given in the stem with an understanding of the SBO/ELAP attachments for repowering critical buses and a knowledge of the preferred lineup for repowering buses.

10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

PPM 5.6.1 Rev: Major: 0 Minor: N/A Page 4 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Comments /

Reference:

PPM 5.6.2 Rev: Major: 15 Minor: N/A Page 5 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Page 6 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-76 Page 7 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-77 Examination Outline Cross-reference: 77 Revision: 2 Date: 11/17/20 Tier: 1 Group: 1 K/A Number: 295004.AA2.03 Level of Difficulty: 3 SRO Importance Rating: 2.9 K/A

Description:

Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Battery voltage CGS is in Mode 1.

Battery Charger E-C1-7 has failed.

ABN-ELEC-125VDC has been entered.

What is the maximum voltage where ABN-ELEC-125VDC requires that Battery E-B1-7 be removed from service?

A. 115 VDC B. 110 VDC C. 105 VDC D. 100 VDC Answer: C K/A Match:

Requires knowledge of battery voltage require to remove battery from service per subsequent actions of ABN-ELEC-125VDC Page 1 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-77 SRO Only:

Page 2 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-77 Explanation:

A. Incorrect. Plausible since it is less than normal DC battery voltage and requires specific knowledge of the ABN-ELEC-125VDC requirement.

B. Incorrect. Plausible since it is less than normal DC battery voltage and requires specific knowledge of the ABN-ELEC-125VDC requirement.

C. Correct. ABN-ELEC-125VDC requires removing the battery from service if BOP125 VDC battery voltage decreases to 105 VDC.

D. Incorrect. Plausible since it is less than normal DC battery voltage and requires specific knowledge of the ABN-ELEC-125VDC requirement. However in this case the voltage is less than the requirement to remove the battery from service.

Technical Reference(s)

ABN-ELEC-125VDC Attached w/ Revision #

See Comments / Reference Proposed references to be provided during examination: None Learning Objective: 15760 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-ELEC-125VDC.

Question Source: #:

(Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires recalling significant parameter requirements for actions taken in Abnormal procedures.

10 CFR Part 55 Content: 55.43 5 Page 3 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-77 Page 4 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78 Examination Outline Cross-reference: 78 Revision: 0 Date: 11/9/20 Tier: 1 Group: 1 K/A Number: 295006.AA2.06 Level of Difficulty: 2 SRO Importance Rating: 3.8 K/A

Description:

Ability to determine and/or interpret the following as they apply to SCRAM: Cause of reactor SCRAM CGS is in Mode 1.

Reactor power is 100%.

Reactor power lowers to 96%

Parameters indicate that a Jet Pump failure has occurred.

A Jet Pump Hold Down Beam failure is confirmed.

What action(s) should the CRS direct?

A. Scram the reactor and secure the affected RRC pump.

B. Reduce RRC flow to 74 Mlbm/hour at 5% per minute.

C. Insert control rods per the Fast Shutdown Sequence.

D. Adjust the unaffected RRC pump speed as necessary to balance loop flows.

Answer: A K/A Match:

Requires knowledge of the requirements to scram the reactor.

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78 SRO Only:

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78 Explanation:

A. Correct. In accordance with ABN-POWER, Unplanned Reactor Power Change, section 4.2.4, if a Jet Pump Hold Down Beam failure is confirmed, then SCRAM the reactor and stop the affected RRC pump.

B. Incorrect. Plausible since this action is required by ABN-POWER, section 4.3, for an unplanned feedwater temperature reduction. However, for the conditions given, reducing core flow may result in entering an unstable area of the Power vs. Flow map.

C. Incorrect. Plausible since this action is required in accordance with TS LCO 3.4.2 and section 4.2.5 of ABN-POWER, for a Jet Sensing Line failure. However, a reactor scram is required for the conditions given.

D. Incorrect. Plausible since a failure of a Jet Pump Hold Down Beam will cause changes in flow in the affected loop. However, RRC pump speed should not change.

Technical Reference(s)

ABN-POWER, Unplanned Reactor Change Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: None Learning Objective: 10312 - Evaluate plant conditions associated with a RRC Flow Control System Failure and determine if a reactor scram is required.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of required actions for a Jet Pump Hold Down Beam Failure 10 CFR Part 55 Content: 55.43 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78 Comments /

Reference:

ABN-POWER Rev: Major: 016 Minor: 003

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-78

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Examination Outline Cross-reference: 79 Revision: 0 Date: 8/24/20 Tier: 1 Group: 1 K/A Number: 295023.AA2.05 Level of Difficulty: 3 SRO Importance Rating: 4.6 K/A

Description:

Ability to determine and/or interpret the following as they apply to REFUELING ACCIDENTS: Entry conditions of emergency plan.

CGS is in Mode 1.

A seismic event occurs.

The event is verified to be an Operating Basis Earthquake (OBE).

The following indications are observed:

Using the reference provided, what Emergency Action Level (EAL) (including the Classification Category), should be declared?

A. Unusual Event, Irradiated Fuel Event B. Unusual Event, Seismic Event C. Alert, Irradiated Fuel Event D. Alert, Hazardous Event Affecting Safety Systems Page 1 of 12

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Answer: C K/A Match:

Requires interpreting emergency plan entry conditions for a fuel (spent fuel pool) accident and determining emergency classification.

Page 2 of 12

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 SRO Only:

Page 3 of 12

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Explanation:

A. Incorrect. Plausible since SFP level is below 22.3 ft and ARM-RIS-1 and ARM-RIS-2 are in alarm.

However, since SFP is below 10 ft, the EAL is ALERT.

B. Incorrect. Plausible since an OBE has been verified, which is an Unusual Event. However, since SFP is below 10 ft, the EAL is ALERT.

C. Correct. In accordance with PPM 13.1.1, EAL Charts, if SFP level lowers to 10 ft., an ALERT is declared (RA2.3). Escalation criteria is if SFP level lowers to below 0.5 ft.

D. Incorrect. Plausible since a seismic event has occurred, which is listed in 13.1.1, Table 8, Hazardous Events. Plausibility is enhanced since both Fuel Pool pumps will trip off with a low level in the skimmer surge tanks. This will occur well above the level depicted in the question stem. If the Fuel Pool Cooling system is considered a Safety System, then this EAL/Classification is correct. However, Safety Systems are listed in 13.1.1, Table 5. This table does not include the Fuel Pool system.

Technical Reference(s)

PPM 13.1.1, Classifying the Emergency (EAL Charts)

Attached w/ Revision #

PPM 13.1.1a, Classifying the Emergency - Technical Basis See Comments / Reference SD000202, Fuel Pool Cooling System Description Tech Ref 4 Proposed references to be provided during examination: Portions of PPM 13.1.1 Learning Objective: 13463 - State when an emergency classification must be made.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the use of EAL Charts along with understanding of the EAL level that must be declared when multiple EALs are applicable.

10 CFR Part 55 Content: 55.43 1 Page 4 of 12

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Comments /

Reference:

PPM 13.1.1 Rev: Major: 49 Minor: 1 Page 5 of 12

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Page 6 of 12

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Comments /

Reference:

PPM 13.1.1a Rev: Major: 34 Minor: 1 Page 7 of 12

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Page 8 of 12

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Comments /

Reference:

SD000202 Rev: Major: 15 Minor: 3 Page 9 of 12

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Comments /

Reference:

Proposed Reference Provided Page 10 of 12

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-79 Page 12 of 12

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 Examination Outline Cross-reference: 80 Revision: 0 Date: 11/12/20 Tier: 1 Group: 1 K/A Number: 295025.2.4.4 Level of Difficulty: 3 SRO Importance Rating: 4.7 K/A

Description:

Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

CGS is in Mode 1.

Reactor power is 100%.

DEH is in automatic.

RPV pressure is 1038 psig, up slow.

What actions should the CRS take?

The CRS should enter (1) and direct lowering pressure using DEH in (2) .

A. (1) PPM 5.1.1, RPV Control (2) BPV MANUAL B. (1) PPM 5.1.1, RPV Control (2) TP MANUAL C. (1) ABN-PRESSURE, Unplanned Reactor Pressure Change (2) BPV MANUAL D. (1) ABN-PRESSURE, Unplanned Reactor Pressure Change (2) TP MANUAL Answer: D K/A Match:

Requires knowledge of the entry conditions for ABN-PRESSURE and PPM 5.1.1.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 Explanation:

A. Incorrect. Plausible since PPM 5.1.1 is entered when RPV pressure is high (1060 psig) and pressure may be controlled with DEH BPV MANUAL when in PPM 5.1.1. However, for the conditions given, ABN-PRESSURE is entered. Additionally, BPV MANUAL is only used if TP MANUAL is not effective in lowering pressure.

B. Incorrect. Plausible since, for the conditions given, pressure is controlled with DEH in TP MANUAL. However, for the conditions given, ABN-PRESSURE is entered.

C. Incorrect. Plausible since ABN-PRESSURE is entered for the conditions given. However, BPV MANUAL is only used if TP MANUAL is not effective in lowering pressure.

D. Correct. ABN-PRESSURE is entered based on an unplanned RPV pressure change. PPM 5.1.1 is only entered if RPV pressure is approaching 1060 psig. In accordance with ABN-PRESSURE, section 4.2, Pressure control is first attempted with DEH in TP MANUAL. If this is not effective in stemming the pressure rise, DEH is placed in BPV MANUAL.

Technical Reference(s)

ABN-PRESSURE, Unplanned Reactor Pressure Change Attached w/ Revision #

PPM 5.1.1, RPV Control See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 15788 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-PRESSURE.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information given in the question with an understanding of EOP and ABN entry requirements, along with a knowledge of ABN subsequent actions to lower pressure.

10 CFR Part 55 Content: 55.43 5 Page 3 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 Comments /

Reference:

ABN-PRESSURE Rev: Major: 015 Minor: 002 Page 4 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 Page 5 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-80 Comments /

Reference:

PPM 5.1.1 Rev: Major: 028 Minor: N/A Page 6 of 6

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 Examination Outline Cross-reference: 81 Revision: 0 Date: 11/12/20 Tier: 1 Group: 1 K/A Number: 295028.EA2.02 Level of Difficulty: 2 SRO Importance Rating: 3.9 K/A

Description:

Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE:

Reactor pressure CGS is in Mode 1.

An event causes the crew to enter PPM 5.2.1, Primary Containment Control.

Concerning HCTL, what action is used to combat exceeding the limit?

A. Raising Wetwell level.

B. Lowering RPV pressure.

C. Lowering Wetwell pressure.

D. Raising Wetwell temperature.

Answer: B K/A Match:

Requires knowledge of the strategy of controlling Reactor Pressure to maintain primary containment integrity with a high Wetwell temperature.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 Explanation:

A. Incorrect. Plausible since Wetwell level is a component of the HCTL curve and a higher Wetwell level gives more margin to HCTL. However, changing Wetwell level to prevent exceeding HCTL is not a strategy used in the EOPs.

B. Correct. In accordance with PPM 5.2.1, Primary Containment Control, step WT-4, when parameters cannot be maintained below HCTL, an emergency depressurization is performed to lower RPV pressure.

C. Incorrect. Plausible since lowering Wetwell pressure will lower Wetwell temperature and increase the margin to HCTL. However, this is not the strategy employed by PPM 5.2.1.

D. Incorrect. Plausible Wetwell temperature is a component of HCTL. However, raising temperature will reduce the margin to HCTL.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control Attached w/ Revision #

PPM 5.0.10, Flowchart Training Manual See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: None Learning Objective: 8303 - Given a list, identify the statement that describes the reason for emergency depressurizing the RPV if wetwell temperature and reactor pressure cannot be maintained below the HCTL.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must synthesize information given in the stem with a knowledge of the parameters that affect HCTL along with the actions required by the EOPs when HCTL is exceeded.

10 CFR Part 55 Content: 55.43 5 Page 3 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-81 Comments /

Reference:

PPM 5.2.1 Rev: 28 Comments /

Reference:

PPM 5.0.10 Rev: Major: 022 Minor: 000 Page 4 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Examination Outline Cross-reference: 82 Revision: 0 Date: 11/16/20 Tier: 1 Group: 1 K/A Number: 295030.2.4.1 Level of Difficulty: 3 SRO Importance Rating: 4.8 K/A

Description:

Knowledge of EOP entry conditions and immediate action steps.

CGS is in Mode 2.

A reactor plant startup is in progress.

Reactor power is 7%.

RPV pressure is 600 psig.

The following annunciators are in alarm:

4.601.A11.2-3: SUPPRESSION POOL LEVEL HIGH/LOW

4.601.A12.2-3: SUPPRESSION POOL LEVEL HIGH/LOW Suppression Pool level:

Suppression Pool level is down slow.

The crew is taking actions in accordance with the ARPs for the annunciators in alarm.

What actions should be taken and what is the minimum Suppression Pool level required to continue unrestricted plant operation?

Entry into PPM 5.2.1, Primary Containment Control (1) required. Suppression Pool level must be controlled GT a minimum of (2) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

A. (1) IS (2) -2.00 inches Narrow Range Page 1 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 B. (1) IS (2) 19.17 feet Wide Range C. (1) IS NOT (2) -2.00 inches Narrow Range D. (1) IS NOT (2) 19.17 feet Wide Range Answer: A K/A Match:

Requires knowledge of the entry conditions for PPM 5.2.1, Primary Containment Control.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Explanation:

A. Correct. Entry into PPM 5.2.1 is required by the ARP even though the plant is in Mode 3.

Additionally, the crew should enter LCO 3.3.6.2, Suppression Pool Water Level since Suppression Pool Level is below 30 ft 9.75 inches or 30.81 inches Wide Range (-2 inches Narrow Range). Condition A requires Suppression Pool level to be restored to GT the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. Incorrect. Plausible since entry into PPM 5.2.1 is required. Plausibility is enhanced since 19.17 feet (19 feet, 2 inches) is the minimum level required to be maintained without performing an emergency depressurization. However, for unrestricted operation to continue, Suppression Pool level must be restored to GT -2 inches Narrow Range in accordance with LCO 3.6.2.2, condition A.

C. Incorrect. Plausible since Suppression Pool level must be restored GT -2.00 inches Narrow Range within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Plausibility is enhanced if it is believed that PPM 5.2.1 is not required to be entered in Mode 3. However, PPM 5.2.1 should be entered regardless of Mode.

D. Plausible if it is believed that PPM 5.2.1 is not required to be entered in Mode 3. Plausibility is enhanced since 19.17 feet (19 feet, 2 inches) is the minimum level required to be maintained without performing an emergency depressurization. However, PPM 5.2.1 should be entered regardless of Mode and LCO 3.6.2.2, condition A requires Suppression Pool level to be restored to GT the limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months (30.81 feet Wide Range or -2.00 inches Narrow Range).

Technical Reference(s)

ARP for 4.601.A11.2-3 (same as 4.601.A12.2-3)

Attached w/ Revision #

PPM 5.1.1, Primary Containment Control See Comments / Reference LCO 3.6.2.2, Suppression Pool Water Level Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 5643 - Referencing Columbia Generating Station Technical Specifications (section 3 only for initial license candidates) associated with the Primary Containment System and a set of plant conditions; determine as applicable the LCO, the action statement and the appropriate bases.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Page 4 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Justification for Cognitive Level Requires Candidate to synthesize information given in the question with a knowledge of TS action statements along with requirements for entry into PPM 5.2.1.

10 CFR Part 55 Content: 55.43 2 Comments /

Reference:

ARP 4.601.A11.2-3 Rev: Major: 026 Minor: N/A Page 5 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Page 6 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A Comments /

Reference:

LCO 3.6.2.2 Rev: Major: 258 Minor: N/A Page 7 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-82 Page 8 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-83 Examination Outline Cross-reference: 83 Revision: 0 Date: 8/26/20 Tier: 1 Group: 2 K/A Number: 295008.2.2.38 Level of Difficulty: 3 SRO Importance Rating: 4.5 K/A

Description:

High Reactor Water Level: Knowledge of conditions and limitations in the facility license.

CGS is in Mode 1.

Reactor power is 50%

The Feedwater Level Control System has failed, providing maximum feed.

In accordance with Technical Specifications, what is the thermal safety limit of concern and how does the plant respond to prevent exceeding this limit?

The challenge to (1) is mitigated by an automatic (2) .

A. (1) Minimum Critical Power Ratio (MCPR)

(2) reactor scram B. (1) Minimum Critical Power Ratio (MCPR)

(2) trip of both Feedwater Pumps C. (1) Linear Heat Generation Rate (LHGR)

(2) reactor scram D. (1) Linear Heat Generation Rate (LHGR)

(2) trip of both Feedwater Pumps Answer: A K/A Match:

Requires knowledge of the effects of high reactor water level on the facility license (technical specificatons).

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-83 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-83 Explanation:

A. Correct. In accordance with the technical specification basis for TS LCO 3.3.2.2, Feedwater and Main Turbine High Water Level Trip Instrumentation, a reactor scram is indirectly caused by RPV level going greater than Level 8 due to a main turbine trip (with reactor power GE 29.5%). Since the question stem states that reactor power is 50%, this trip will occur. Additionally, the basis states that the reactor scram mitigates the reduction in MCPR.

B. Incorrect. Plausible since MCPR is the limit of concern. Plausibility is enhanced since a trip of both feedwater pumps is automatically inserted when RPV level reaches Level 8. However, the automatic main turbine trip and the associated reactor scram mitigates the reduction in MCPR.

C. Incorrect. Plausible since the main turbine trip and subsequent reactor scram mitigates the challenge to the limit of concern. However, this limit is MCPR.

D. Incorrect. Plausible since a trip of both feedwater pumps is automatically inserted when RPV level reaches Level 8. However, the limit of concern is MCPR and a main turbine trip and subsequent reactor scram mitigates the challenge to this limit.

Technical Reference(s)

Technical Specification Bases Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 13425 - Describe the bases for the Minimum Critical Power Ratio Safety Limit.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must demonstrate an understanding of the challenge to reactor safety from a high reactor water level event along with a knowledge of the automatic actions that mitigate this challenge.

10 CFR Part 55 Content: 55.43 2 Page 3 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-83 Comments /

Reference:

Technical Specification Bases Rev: Major: 98 Minor: N/A Page 4 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 Examination Outline Cross-reference: 84 Revision: 0 Date: 8/26/20 Tier: 1 Group: 2 K/A Number: 295033.2.4.21 Level of Difficulty: 3 SRO Importance Rating: 4.6 K/A

Description:

High Secondary Containment Area Radiation Levels: 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.

CGS is in Mode 1.

An event occurs which causes high secondary containment radiation levels.

The crew enters PPM 5.3.1, Secondary Containment Control.

The CRS is evaluating secondary containment area radiation levels against their Maximum Safe Operating Value (MSOV).

What is the basis for the MSOV value?

The MSOV is high enough to (1) and low enough to allow time for shutdown or leak isolation without exceeding (2) .

A. (1) prevent MSOV violations during normal plant operation (2) operator normal dose limits during event response B. (1) prevent MSOV violations during normal plant operation (2) the allowable dose of the most sensitive safety related equipment C. (1) indicate a substantial and immediate problem (2) operator normal dose limits during event response D. (1) indicate a substantial and immediate problem (2) the allowable dose of the most sensitive safety related equipment Answer: D K/A Match:

Requires knowledge of logic used to assess secondary containment radiation levels to determine containment conditions.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 Explanation:

A. Incorrect. Plausible since the MSOV is much higher than normal radiation levels. Plausibility is enhanced since all actions are designed to minimize dose. However, the MSOV is set high enough to alert operators that there is a condition that represents a substantial risk to elevated dose to the public and requires immediate action.

B. Incorrect. Plausible since the MSOV is set low enough to prevent exceeding the total integrated dose allowable for the most sensitive safety related equipment. However, , the MSOV is set high enough to alert operators that there is a condition that represents a substantial risk to elevated dose to the public and requires immediate action.

C. Incorrect. Plausible since the MSOV is set high enough to alert operators that there is a condition that represents a substantial risk to elevated dose to the public and requires immediate action.

However, the MSOV is set low enough to prevent exceeding the total integrated dose allowable for the most sensitive safety related equipment.

D. Correct. PPM 5.0.10, Flowchart Training Manual, specifies that the MSOV for secondary containment radiation levels is high enough to be indicative of substantial and immediate problems yet low enough to allow time for shutdown or isolation of a leak without exceeding the total integrated dose allowable for even the most sensitive safety related equipment.

Technical Reference(s)

PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8456 - Define Maximum Safe Operating Value for the following secondary containment parameters: c. Area radiation levels Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the reasons for the MSOV value.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-84 10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

PPM 5.0.10 Rev: Major: 022 Minor: 000 Page 4 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 Examination Outline Cross-reference: 85 Revision: 1 Date: 11/17/20 Tier: 1 Group: 2 K/A Number: 295036.EA2.02 Level of Difficulty: 3 SRO Importance Rating: 3.1 K/A

Description:

Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Water level in the affected area.

CGS is in Mode 1.

The CRS enters PPM 5.3.1, Secondary Containment Control, due to a primary system discharging into the Reactor Building.

HPCS Pump Room water level is above the alarm setpoint.

If time does not permit removal of floor plugs to determine room water level, how does the crew determine if HPCS Pump Room water level is above its Maximum Safe Operating Value (MSOV)?

HPCS Pump Room water level considered to be above MSOV if A. level is above the alarm setpoint ONLY.

B. level cannot be restored and maintained below the alarm setpoint.

C. there is indication of water weeping from the bottom of the HPCS Room door.

D. level is above the alarm setpoint and the HPCS Room door cannot be opened.

Answer: D K/A Match:

Requires knowledge of alternate method of determining water level in the HPCS Pump Room.

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 SRO Only:

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 Explanation:

A. Incorrect. Plausible since water level above the alarm setpoint is a decision point in PPM 5.3.1, Secondary Containment Control (step SC-6). However, water level is not considered to exceed MSOV unless the room cannot be entered, and the level is above the alarm setpoint.

B. Incorrect. Plausible since the inability to lower room level to below the alarm setpoint indicates that level is still rising and could exceed MSOV. Additionally the inability to lower level below the alarm setpoint is a decision point in PPM 5.3.1 (step SC-8) that determines if primary system isolations should be attempted. However, determining that room level is above MSOV without removing floor plugs requires that the room door cannot be opened.

C. Incorrect. Plausible since water level high enough to leak past the door could indicate that room level is greater than MSOV. However, HPCS Pump Room is determined to be greater than MSOV at 69 inches above the floor, which would seal the door and not allow any water to leak.

D. Correct. In accordance with PPM 5.5.27, Reactor Building 422 Max Safe Operating Level Measurement, if the affected room door cannot be opened, and time does not permit removal of floor plugs, then consider the level in the room is above the Max Safe Operating Level.

Technical Reference(s)

PPM 5.3.1, Secondary Containment Control Attached w/ Revision #

PPM 5.5.27, RB 422 Max Safe Operating Level Measurement See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 69220 - Given plant conditions identify those symptoms that would indicate Reactor Building 422 Area Flooding.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of the method to determine if room level is greater than MSOV when floor plugs cannot be removed.

10 CFR Part 55 Content: 55.43 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 Comments /

Reference:

PPM 5.3.1 Rev: 21

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-85 Comments /

Reference:

PPM 5.5.27 Rev: Major: 005 Minor: N/A

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Examination Outline Cross-reference: 86 Revision: 1 Date: 7/8/20 Tier: 2 Group: 1 K/A Number: 209001.2.2.12 Level of Difficulty: 3 SRO Importance Rating: 4.1 K/A

Description:

Low Pressure Core Spray: Knowledge of surveillance procedures.

CGS is in Mode 1.

Surveillance OSP-LPCS/IST-Q702, LPCS System Operability Test, has just been completed.

The surveillance was performed as an ASME required Comprehensive Pump Test (CPT).

Partial surveillance results for LPCS-P-1:

Suction Pressure (Fluke Module): 15.0 psig

Discharge Pressure (Fluke Module): 310.0 psig

Differential pressure p (Discharge Pressure-Suction Pressure): 295 psig

Indicated flowrate (TDAS X164): 6410 gpm Surveillance results for LPCS-V-12:

Opening time: 22.32 seconds

Closing time: 14.74 seconds Using the references provided, what is the next action that is required to be performed?

A. Perform a second stroke time test of LPCS-V-12 immediately.

B. Close and de-activate LPCS-V-12 within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

C. Restore LPCS to operable status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

D. Restore LPCS to operable status within 7 days.

Answer: D K/A Match:

Requires a knowledge of LPCS System Operability surveillance and its relationship to technical specification requirements.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Explanation:

A. Incorrect. Plausible since the raw closing stroke time of LPCS-V-12 is outside the Alert Level in accordance with OSP-LPCS/IST-Q702, Attachment 9.1. Section 4.6.2 states that valves with measurements outside the Alert Value shall be immediately retested or declared inoperable.

However, note (+2) or Attachment 9.1 gives guidance for rounding times to the nearest second, which will put LPCS-V-12 within the Alert Level.

B. Incorrect. Plausible since the raw closing stroke time of LPCS-V-12 is outside the Alert Level in accordance with OSP-LPCS/IST-Q702, Attachment 9.1. Section 4.6.2 states that valves with measurements outside the Alert Value shall be immediately retested or declared inoperable. If declared inoperable, LPCS-V-12 would need to be closed and deenergized in accordance with LCO 3.6.1.3. However, note (+2) or Attachment 9.1 gives guidance for rounding times to the nearest second, which will put LPCS-V-12 within the Alert Level.

C. Incorrect. Based on the surveillance results given, p is in the Action Range on Attachment 9.8.

Section 4.7 states that the LPCS pump must be declared inoperable. If LCO 3.5.1.C is interpreted as one ECCS spray system inoperable then 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months to restore to operable could be interpreted as the right answer.

D. Correct. Based on the surveillance results given, p is in the Action Range on Attachment 9.8.

Section 4.7 states that the LPCS pump must be declared inoperable. This will make LCO 3.5.1.A applicable, which requires the LPCS system to be in operable status within 7 days.

Technical Reference(s)

OSP-LPCS/IST-Q702, LPCS System Operability Test Attached w/ Revision #

TS LCO 3.5.1, ECCS Operating See Comments / Reference Proposed references to be provided during examination: OSP-LPCS/IST-Q702 (Partial)

TS LCO 3.5.1 (Partial)

TS LCO 3.6.1.3 LCS Table 1.6.1.3 (Partial)

Learning Objective: 5487 - Referencing Columbia Generating Station Technical Specifications associated with the Low Pressure Core Spray System and a set of plant conditions, determine as applicable the LSSS, the LCO, the action statement, and the appropriate bases.

Question Source: #:

(Note changes or attach parent)

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize the conditions listed in the question using a knowledge of TS LCO 3.5.1 and an understanding to the LPCS System Operability surveillance.

10 CFR Part 55 Content: 55.43 2 Comments /

Reference:

OSP-LPCS/IST-Q702 Rev: Major: 044 Minor: 001 Page 4 of 13

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-86 Comments /

Reference:

LCO 3.5.1 Rev: Major: 258 Minor: N/A Page 10 of 13

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-87 Examination Outline Cross-reference: 87 Revision: 0 Date: 8/27/20 Tier: 2 Group: 1 K/A Number: 211000.2.1.31 Level of Difficulty: 2 SRO Importance Rating: 4.3 K/A

Description:

Standby Liquid Control: Ability to locate control room switches, controls, and indications, and to determine that they correctly reflect the desired plant lineup.

CGS is in Mode 1 The following annunciator is in alarm:

4.603.A8.6-8: SLC DIV 2 OUT OF SERVICE SLC system indications:

SLC-V-4B LOSS OF CONTINUITY How does this affect the operation and technical specification operability of SLC?

SLC train B (1) operable. If initiated, SLC-P-1B (2) inject into the RPV.

A. (1) is (2) will B. (1) is (2) will not C. (1) is not (2) will D. (1) is not (2) will not Answer: C K/A Match:

Requires understanding of SLC abnormal indications and effect on plant operations.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-87 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-87 Explanation:

A. Incorrect. Plausible since SLC-P-1B will inject to the RPV via SLC-V-4A. Plausiblity is enhanced since the SLC system will still provide 100% flow. However, SLC Train B is considered inoperable.

B. Incorrect. Plausible since the SLC system will still provide 100% flow. Plausibility is enhanced since the Train B discharge squib valve, SLC-V-4B, will not open on an initiation. However, on initiation, SLC-P-1B will inject to the RPV via SLC-V-4A. Additionally, SLC Train B is considered inoperable.

C. Correct. In accordance with Technical Specification Bases for LCO 3.7.1, to be considered operable, each SLC train must have an operable pump, explosive valve and associated piping, valves and instruments and controls. Since the SLC pump discharge paths are cross-connected upstream of the squib valves, SLC-P-1B will still inject to the RPV on SLC initiation.

D. Incorrect. Plausible since SLC Train B is considered inoperable. Plausibility is enhanced since the Train B discharge squib valve, SLC-V-4B, will not open on an initiation. However, on initiation, SLC-P-1B will inject to the RPV via SLC-V-4A.

Technical Reference(s)

Technical Specification Basis for LCO 3.7.1, SLC System Attached w/ Revision #

SD000172, Standby Liquid Control System Description See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7636 - Explain how the SLC Explosive valve detonator continuity is verified.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires an understanding of the operation of the SLC system with a single squib valve failed closed along with a knowledge of the operability requirements for individual trains of SLC.

10 CFR Part 55 Content: 55.43 2 Page 3 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-87 Comments /

Reference:

TS Basis for LCO 3.7.1 Rev: Major: 87 Minor: N/A Page 4 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-87 Comments /

Reference:

SD000172 Rev: Major: 13 Minor: 1 Page 5 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 Examination Outline Cross-reference: 88 Revision: 0 Date: 11/16/20 Tier: 2 Group: 1 K/A Number: 261000.A2.09 Level of Difficulty: 3 SRO Importance Rating: 2.6 K/A

Description:

Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Plant air system failure CGS is in Mode 1.

A complete Loss of CAS occurs.

How should the CRS direct controlling Reactor Building (RB) pressure?

The CRS should direct operators to...

A. verify RB pressure is being controlled LTE -0.25 WG in accordance with SOP-HVAC/RB-OPS, Reactor Building Ventilation System Operation.

B. secure one train of SGT in accordance with SOP-SGT-OPS, since both SGT trains automatically started.

C. restart RB HVAC in accordance with SOP-RBHVAC-RESTART-QC, since both trains of SGT are isolated due to the loss of CAS.

D. start one train of SGT in accordance with SOP-SGT-START-DIV1(2)-QC since SGT trains did not isolate on the loss of CAS.

Answer: D K/A Match:

Requires knowledge of the effects on SGT with a loss of CAS and a knowledge of the procedure used to maintain Reactor Building pressure.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 Explanation:

A. Incorrect. Plausible since RB HVAC fans do not trip directly from a loss of CAS. Plausibility is enhanced if it is believed that SGT-V-2A(B), Reactor Building Intake Isolation Valves, fail closed on a loss of CAS. However, REA/ROA dampers fail closed on a loss of CAS. This causes the REA/ROA fans to trip on differential pressure.

B. Incorrect. Plausible since one train of SGT should be secured per SOP-SGT-OPS, if both trains are operating. Plausibility is enhanced if it is believed that SGT auto-starts on a loss of CAS.

However, SGT does not auto-start and must be manually started.

C. Incorrect. Plausible PPM 5.3.1, step SC-1 directs re-starting RB HVAC if SGT cannot restore RB differential pressure. Plausibility is enhanced since SGT-V-2A and 2B use CAS. However, SGT-V-2A and 2B fail open, which is the position necessary to operate SGT. Therefore one train of SGT is started.

D. Correct. In accordance with ABN-CAS, Control Air System Failure, a complete loss of CAS will cause REA and ROA isolation valves to fail closed, and RB HVAC will be isolated. Step 4.6 directs operators to restore RB differential pressure by starting one train of SGT per SOP-SGT-START-DIV/1(2)- QC, or SOP-SGT-START. The note prior to step 4.6 informs operators that SGT-V-2A(B) fail open on a loss of CAS and do not need to be manually repositioned.

Technical Reference(s)

ABN-CAS, Control Air System Failure Attached w/ Revision #

PPM 5.3.1, Secondary Containment Control See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 15734 - With the procedures available, discuss all contingencies associated with the subsequent operator actions of ABN-CAS.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires candidate to synthesize the information in the question with a knowledge of system response to a loss of CAS along with an understanding of ABN-CAS supplemental actions.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

ABN-CAS Rev: Major: 011 Minor: 001 Page 4 of 7

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-88 Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor: N/A Page 7 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Examination Outline Cross-reference: 89 Revision: 0 Date: 11/13/20 Tier: 2 Group: 1 K/A Number: 263000.A2.02 Level of Difficulty: 3 SRO Importance Rating: 3.0 K/A

Description:

Ability to (a) predict the impacts of the following on the D.C. ELECTRICAL DISTRIBUTION; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of ventilation during charging CGS is in Mode 1.

Div. 1 125 VDC battery is in equalize.

A loss of Div. 1 125 VDC Battery Room ventilation occurs.

Div. 1 125 VDC Battery Room temperature is 72°F.

Using the reference provided, what is the earliest action required by the Licensee Controlled Specifications (LCS) and why is the action performed?

Verify (1) . This action is performed to (2) .

A. (1) battery is not on equalize (2) prevent accumulation of hydrogen gas B. (1) battery is not on equalize (2) ensure requirements for station blackout are met C. (1) battery room temperature is within requirements (2) prevent accumulation of hydrogen gas D. (1) battery room temperature is within requirements (2) ensure requirements for station blackout are met Answer: D K/A Match:

Requires knowledge of actions to take if battery room ventilation fails and why those actions are taken.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Explanation:

A. Incorrect. Plausible since verification that the battery is not in equalize is a requirement for the conditions given in the stem and the reason for this action is correct. However, LCS 1.8.6.2 requires that the room temperature be verified within limits first.

B. Incorrect. Plausible since verification that the battery is not in equalize is a requirement for the conditions given in the stem. Plausibility is enhanced since the reason given in the distractor is the correct reason for temperature verification. However, LCS 1.8.6.2 requires that the room temperature be verified within limits first.

C. Incorrect. Plausible since verifying room temperature within requirements is the first action that should be performed in accordance with LCS 1.8.6.2. Plausibility is enhanced since the reason for the action is correct for verification that the battery is not in equalize. However, the reason for verifying temperature is to ensure that battery performance will meet station blackout requirements.

D. Correct. In accordance with LCS 1.8.6.2, if battery room ventilation is lost, room temperature is verified to meet requirements within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The battery may remain in equalize for up to 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . This action is taken to ensure that the battery will meet station blackout requirements.

Technical Reference(s)

LCS 1.8.6.2, 125 and 250 VDC Battery Parameters Attached w/ Revision #

LCS 1.8.6.2 Bases See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: LCS 1.8.6.2 Learning Objective: 6925 - Identify the basis for any LCO.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: Exam year or N/A Question Cognitive Level:

Justification for Cognitive Level Page 3 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Requires the examinee to synthesize conditions given in the question stem using a knowledge of LCS compensatory actions along with a knowledge of the reasons for these actions.

10 CFR Part 55 Content: 55.43 2 Comments /

Reference:

LCS 1.8.6.2 Rev: Major: 114 Minor: N/A Page 4 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Comments /

Reference:

LCS 1.8.6.2 Bases Rev: Major: 114 Minor: N/A Page 5 of 9

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-89 Comments /

Reference:

Reference Provided to Candidates Rev: Major: Maj Minor: Min Page 6 of 9

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Examination Outline Cross-reference: 90 Revision: 0 Date: 11/13/20 Tier: 2 Group: 1 K/A Number: 400000.A2.01 Level of Difficulty: 2 SRO Importance Rating: 3.4 K/A

Description:

Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Loss of CCW pump CGS is in Mode 1.

RCC-P-1A and RCC-P-1C are running.

The supply breaker to SL-81, CB-8/81, trips.

How is RCC affected and what action(s) should the CRS direct?

There is a (1) . Direct control room operators to (2) .

A. (1) partial loss of RCC flow (2) manually start the standby RCC pump B. (1) complete loss of RCC flow to the Drywell (2) manually SCRAM the reactor C. (1) partial loss of RCC flow (2) place RCC-P-1B and RCC-P-1C in PTL.

D. (1) complete loss of RCC flow to the Radwaste/Rx Building (2) verify that the standby RCC pump has automatically started Answer: C K/A Match:

Requires knowledge of the effects of losing power to a running RCC pump and the actions required.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Explanation:

A. Incorrect. Plausible since a partial loss of RCC flow has occurred. Plausibility is enhanced since the standby RCC pump will not automatically start on a loss of one RCC pump when the lost pumps breaker is still closed (ie - shaft shear, loss of power to the bus, etc.) and ABN-RCC, step 4.2.1 requires operators to manually start the standby RCC pump if not running. However, since RCC-P-1B is also powered from SL-81, it cannot be manually started.

B. Incorrect. Plausible since the RCC Outboard Supply valve (RCC-V-104) and Inboard Return valve (RCC-V-40) are powered from MC-8B-A, which loses power when CB-8/81 trips, and a candidate may believe that these valves will close. Plausibility is enhanced since ABN-RCC, step 4.1, requires a manual scram on a complete loss of RCC flow to the drywell. However, the drywell isolation valves do not reposition on a loss of power. Therefore, a manual scram is not required.

C. Correct. RCC-P-1B and RCC-P-1C are powered from SL-81 while RCC-P-1A is powered from SL-71. Normally, the standby RCC pump will automatically start on a loss of a running RCC pump (less than two RCC pump breakers closed). However, for the conditions given, RCC-P-1C breaker remains closed, so RCC-P-1B will not receive a start signal. Even if a start signal was received, RCC-P-1B could not start since it has no power. RCC-V-6 automatically closes when less than two RCC pump breakers are closed for greater than 10 seconds. It closes to ensure that sufficient RCC flow is directed to vital equipment in the drywell. For the conditions given, RCC-V-6 will not automatically close. On a loss of SL-81, ABN-ELEC-SM3/SM8, step 4.5.1, requires operators to place RCC-P-1B and RCC-P-1C to PTL. This will open the pump breakers and cause RCC-V-6 to automatically close.

D. Incorrect. Plausible since a complete loss of flow to the Radwaste/Rx Building should have occurred with only one RCC pump running (see explanation C above), and ABN-RCC, step 4.2.1, requires operators to verify that the standby RCC pump has started. However, since all RCC pump breakers are closed and RCC-P-1B is not powered, the standby RCC pump will not start and cannot be started manually. Additionally, RCC-V-6 will not automatically close.

Technical Reference(s)

SD000196, RCC ABN-RCC, Loss of RCC Attached w/ Revision #

ABN-ELEC-SM3/SM8, SM-3, SM-8, SM-85, SM-82, SL-81, SL-83 & See Comments / Reference SL-31 Distribution System Failures Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 7668 - Predict the plant response to a: a. Partial and a complete loss of the RCC system.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee synthesize information given in the question with an understanding of the requirements for automatic actions of the RCC system along with procedural requirements of ABN-RCC and ABN-ELEC-SM3/SM8 10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

SD000196 Rev: Major: 14 Minor: 2 Page 4 of 7

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Comments /

Reference:

ABN-RCC Rev: Major: 006 Minor: 004 Page 6 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-90 Comments /

Reference:

ABN-ELEC-SM3/SM8 Rev: Major: 022 Minor: N/A Page 7 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Examination Outline Cross-reference: 91 Revision: 0 Date: 11/17/20 Tier: 2 Group: 2 K/A Number: 216000.2.2.22 Level of Difficulty: 3 SRO Importance Rating: 4.7 K/A

Description:

Nuclear Boiler Instrumentation - Knowledge of limiting conditions for operations and safety limits.

CGS is in Mode 1.

A mechanical failure of MS-PS-23A (RPS High RPV Pressure Instrument) occurs.

The CRS enters TS LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation.

All required TS actions for this condition are complete.

15 hours1.736111e-4 days 0.00417 hours 2.480159e-5 weeks 5.7075e-6 months later the crew performs OSP-INST-H101, Shift and Daily Instrument Checks (Modes 1, 2, 3).

MS-PS-23D (RPS High RPV Pressure Instrument), fails its channel check.

Using the reference provided, what is the most limiting TS action that should be taken for this condition?

A. Restore RPS trip capability in 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

B. Place a channel in one trip system in trip in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

C. Place MS-PS-23D in trip in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D. Be in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Answer: D K/A Match:

Requires knowledge of the LCOs for the Nuclear Boiler Instrumentation system.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Explanation:

A. Incorrect. Plausible since two channels are inoperable. However, When channel A failed, the question states that appropriate TS actions were taken, which means that action A1 was performed and channel A was placed in TRIP. Therefore, trip capability is maintained.

B. Incorrect. Plausible since one channel is inoperable in both trip systems. However, channel A is already in trip. Therefore, required action B.1 is complete.

C. Incorrect. Plausible since required action A.1 should be completed for channel D. However, this will cause a reactor scram. LCO 3.3.1.1 bases states Alternately, if it is not desired to place the channel (or trip system) in trip (e.g., as in the case where placing the inoperable channel in trip would result in a full scram), Condition D must be entered and its Required Action taken.

D. Correct. With a failure of MS-PS-23D, LCO 3.3.1.1 required action A.1 requires the channel to be placed in Trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . However, LCO 3.3.1.1 bases states that if this action will cause a scram, condition D should be entered, which requires the reactor to be placed in Mode 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Technical Reference(s)

LCO 3.3.1.1, Reactor Protection System (RPS) Instrumentation Attached w/ Revision #

LCO 3.3.1.1 Bases See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: LCO 3.3.1.1 Actions and Table 3.3.1.1-1 Learning Objective: 11776 - Referencing Columbia Generating Station Licensee Controlled Specifications associated with the Nuclear Boiler Instrumentation System and a set of plant conditions, determine as applicable the LCO, the action statement, and the appropriate bases.

Question Source: #: LO03048

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing conditions given in the question with a knowledge of LCO 3.3.1.1 action statements along with the basis for these actions Page 3 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 10 CFR Part 55 Content: 55.43 2 Comments /

Reference:

LCO 3.3.1.1 Rev: Major: 258 Minor: N/A Page 4 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Page 5 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Comments /

Reference:

LCO 3.3.1.1 Bases Rev: Major: 114 Minor: N/A Page 6 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-91 Comments /

Reference:

Provided Reference Rev: Major: 258 Minor: N/A Page 7 of 10

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-92 Examination Outline Cross-reference: 92 Revision: 0 Date: 7/14/20 Tier: 2 Group: 2 K/A Number: 226001.A2.14 Level of Difficulty: 2 SRO Importance Rating: 3.1 K/A

Description:

Ability to (a) predict the impacts of the following on the RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: High suppression pool level.

CGS is in Mode 1.

A LOCA occurs.

The crew is performing actions in accordance with PPM 5.1.1, RPV Control and PPM 5.2.1, Primary Containment Control.

Current plant conditions:

RHR-A is lined up to inject into the RPV.

RHR-B- is spraying the wetwell.

Drywell temperature is 240°F, up slow.

Drywell pressure is 21 psig, up slow.

Wetwell pressure is 15 psig, up slow.

Wetwell level is 52 feet, up slow.

What action should the CRS direct to address Primary Containment parameters and why is this action taken?

The CRS should direct the crew to A. vent Primary Containment in accordance with PPM 5.5.14, Emergency Wetwell Venting, to preclude containment failure.

B. spray the drywell with Service Water in accordance with PPM 5.5.2, RHR/SW Crosstie Lineup, to reduce drywell temperature.

C. emergency depressurize the RPV in accordance with PPM 5.1.3, Emergency RPV Depressurization, to preclude steam in the suppression pool airspace.

D. secure wetwell spray in accordance with SOP-RHR-SPRAY-WW-QC, Initiation of Wetwell Spray - Quick Card, since post-spray drywell vacuum relief may be lost.

Answer: C Page 1 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-92 K/A Match:

Requires knowledge of the actions required when Wetwell level exceeds the Wetwell-to-Drywell level limit.

SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-92 Explanation:

A. Incorrect. Plausible since PPM 5.2.1, step P-14 requires venting primary containment if the Primary Containment Pressure Limit (PCPL) is exceeded. However, PCPL is not exceeded, Pressure Suppression Pressure (PSP) is exceeded which requires an emergency depressurization in accordance with step P-12.

B. Incorrect. Plausible if it is believed that RHR-A is required to inject into the RPV while Drywell spray is required. However, since wetwell level is greater than 51 feet, drywell spray should not be initiated in accordance with PPM 5.2.1, step DT-5. Additionally, Drywell spray is not required to by initiated until drywell temperature approaches 330°F.

C. Correct. In accordance with the information given in the stem, PSP is exceeded. PPM 5.2.1, step P-12 requires the performance of an emergency depressurization.

D. Incorrect. Plausible since wetwell level is greater than 51 feet and PPM 5.2.1, step L-17 requires securing drywell spray due to potential loss of drywell vacuum relief. However, wetwell spray is not required to be secured.

Technical Reference(s)

PPM 5.2.1, Primary Containment Control Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 8301 - Given a list, identify the statement that describes the failure mode that the Heat Capacity Temperature Limit protects against. (PPM 5.2.1).

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must synthesize information given in the question stem with an understanding of flowchart figures (PCPL, PSP, etc.) along with a knowledge of the procedure steps that are required when a limit is exceeded.

10 CFR Part 55 Content: 55.43 5 Page 3 of 8

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-92 Comments /

Reference:

PPM 5.2.1 Rev: Major: 28 Minor: N/A Page 4 of 8

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Examination Outline Cross-reference: 93 Revision: 1 Date: 11/30/20 Tier: 2 Group: 2 K/A Number: 290001.A2.02 Level of Difficulty: 3 SRO Importance Rating: 3.7 K/A

Description:

Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Excessive outleakage CGS is in Mode 1.

Field operators report a breach in the Reactor Building roof.

Plant indications include:

Reactor Building DP is stable.

Which of the following actions should be taken?

1. Enter PPM 5.3.1, Secondary Containment Control, and start one train of SGT.

2. Enter Technical Specification LCO 3.6.4.1, Secondary Containment, and restore secondary containment to operable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Adjust REA-DPIC-1B in MANUAL and restore secondary containment P in accordance with ARP 4.812.R1.7-3, SEC PRESS CONTR A P HIGH/LOW.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 A. 1 ONLY B. 2 ONLY C. 1 and 3 ONLY D. 2 and 3 ONLY Answer: A K/A Match:

Requires knowledge of the consequences of a leak in secondary containment and the procedures that should be entered to mitigate these consequences.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Explanation:

A. Incorrect. Plausible since PPM 5.3.1 is entered on a low vacuum in secondary containment.

Plausibility is enhanced since PPM 5.3.1 contains actions if SGT cannot restore secondary containment P. However, secondary containment P must be 0 inches WG to enter PPM 5.3.1. Additionally, PPM 5.3.1 does not direct starting SGT.

B. Correct. TS 3.6.4.1 requires secondary containment vacuum to be 0.25 inches vacuum WG Additionally. a breach in the Reactor Building roof is also an entry condition for TS 3.6.4.1.

C. Incorrect. Plausible for the reasons listed in explanation A above and since 4.812.R1.7-3 directs operators to place REA-DPIC-1B in MANUAL to restore secondary containment vacuum.

However, this annunciator does not alarm until 0 inches WG. Additionally, REA-DPIC-1B is currently 100% open. Taking the controller to manual will not allow any additional adjust to restore vacuum.

D. Incorrect. Plausible since (2) is correct. However, (3) is not correct. See explanation C above.

Technical Reference(s)

TS 3.6.4.1, Secondary Containment Attached w/ Revision #

PPM 5.3.1, Secondary Containment Control See Comments / Reference 4.812.R1.7-3, Alarm Response OSP-CONT-M102, Secondary Containment Integrity Verification Proposed references to be provided during examination: N/A Learning Objective: 7007 - With the Technical Specifications provided, locate all Safety Limits and/or LCO's that directly relate to the Secondary Containment.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires synthesizing conditions given in the question with a knowledge of procedure entry requirements and required actions.

10 CFR Part 55 Content: 55.43 2 Page 4 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Comments /

Reference:

TS 3.6.4.1 Rev: Major: 254 Page 5 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Page 6 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Page 7 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Comments /

Reference:

OSP-CONT-M102 Rev: Major: 12 Minor: N/A Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor:

Page 8 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Comments /

Reference:

4.812.R1.7-3 Rev: Major: 21 Minor: N/A Page 9 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-93 Page 10 of 10

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-94 Examination Outline Cross-reference: 94 Revision: 1 Date: 11/13/20 Tier: 3 Group: N/A K/A Number: 2.1.7 Level of Difficulty: 3 SRO Importance Rating: 4.7 K/A

Description:

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

CGS is in Mode 2.

A reactor plant startup is in progress in accordance with PPM 3.1.2, Startup Flowchart.

The reactor is critical. The crew is withdrawing control rods to place the reactor in the heating range.

Current plant conditions:

Reactor period is 85 seconds and stable.

All SRMs have been fully withdrawn.

IRMs are on the following ranges:

IRM A: 9 IRM E: 9 IRM B: 10 IRM F: 9 IRM C: 9 IRM G: 9 IRM D: 9 IRM H: 10

The reactor is currently NOT in the heating range.

What action should the CRS direct?

The CRS should direct A. stopping control rod withdrawal and contacting the SNE.

B. withdrawing control rods to achieve a heatup rate of LE 80°F/hr.

C. adjusting control rod position to maintain a stable period of GT 60 seconds.

D. driving control rods in the reverse order until all control rods are fully inserted.

Answer: A K/A Match:

Requires examinee to understand expected reactor plant response when withdrawing control rods during a reactor startup and the correct response when reactor behavior is not as expected.

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-94 SRO Only:

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-94 Explanation:

A. Correct. In accordance with PPM 3.1.2, Startup Flowchart, steps Q24/Q25, if nuclear heating does not occur on or before IRMs on range 8, the CRS should stop rod withdrawal and contact the SNE.

B. Incorrect. Plausible since withdrawing control rods to achieve and maintain a heatup rate would be the next step after the reactor is in the heating range. However, since the reactor is not yet in the heating range, a heatup rate cannot be established.

C. Incorrect. Plausible since the direction to adjust control rods to maintain a period GE 60 seconds is applicable while approaching the heating range. However, since IRM D is on a range > range 8, the CRS should direct that control rod withdrawal be suspended.

D. Incorrect. Plausible since fully inserting control rods is the correct action if the reactor isnt critical within the specified maximum and minimum ECP. However, since the reactor is already critical, this requirement does not apply.

Technical Reference(s)

PPM 3.1.2, Startup Flowchart.

Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 6652 - With the procedures available, determine which steps/sections of the procedure must be repeated if the startup is delayed. [PPM 3.1.2] (SRO only)

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires examinee to synthesize information given in the stem with an understanding of expected reactor response during a reactor startup along with a knowledge of the actions required if reactor response is not as expected.

10 CFR Part 55 Content: 55.43 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-94 Comments /

Reference:

PPM 3.1.2 Rev: Major: 088 Minor: 000

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-94

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-95 Examination Outline Cross-reference: 95 Revision: 0 Date: 11/13/20 Tier: 3 Group: N/A K/A Number: 2.1.35 Level of Difficulty: 2 SRO Importance Rating: 3.9 K/A

Description:

Knowledge of the fuel-handling responsibilities of SROs.

Which of the following evolutions must be directly supervised by a licensed SRO?

(1) Fuel movement from the core to the Spent Fuel Pool.

(2) Movement of LPRMs within the core.

(3) Fuel movement between locations in the core.

(4) Control Rod blade replacement in a defueled cell.

A. (1) and (2) ONLY B. (1) and (3) ONLY C. (2) and (4) ONLY D. (3) and (4) ONLY Answer: B K/A Match:

Requires knowledge of SRO responsibilities during fuel handling evolutions.

SRO Only:

The question requires knowledge of refuel floor SRO responsibilities which is an SRO-only topic as delineated in NUREG-1021, ES-401, Attachment 2 and 10 CFR 55.43(b)(7)

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-95 Explanation:

A. Incorrect. Plausible since (1) is considered a CORE ALTERATION in accordance with TS and must be supervised by a licensed SRO. However, However, (2) is specifically called out in TS as an evolution that is not considered a CORE ALTERATION, and therefore, licensed SRO supervision is not required.

B. Correct. In accordance with TS, both (1) and (3) are considered a CORE ALTERATION, and must be supervised by a licensed SRO.

C. Incorrect. Plausible since (4) must be supervised by a licensed SRO when a control rod blade is replaced in a fueled cell. However, (2) is specifically called out in TS as an evolution that is not considered a CORE ALTERATION and since the control rod blade is being replaced in a non-fueled cell, it is not considered a CORE ALTERATION. Therefore, licensed SRO supervision is not required for either evolution.

D. Incorrect. Plausible since (3) must be supervised by a licensed SRO and 4) must be supervised by a licensed SRO when a control rod blade is replaced in a fueled cell. However, since the control rod blade is being replaced in a non-fueled cell, it is not considered a CORE ALTERATION. Therefore, licensed SRO supervision is not required for this evolution.

Technical Reference(s)

Technical Specifications Attached w/ Revision #

PPM 6.3.2, Fuel Shuffling and/or Offloading and Reloading See Comments / Reference OI-20, Fuel Handling Expectations Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 3051 - Define Core Alteration, including items that are specifically excluded from the definition.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of definition of CORE ALTERATIONS Page 2 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-95 10 CFR Part 55 Content: 55.43 6 Comments /

Reference:

Technical Specifications Rev: Major: 254 Minor: N/A Page 3 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-95 Comments /

Reference:

PPM 6.3.2 Rev: Major: 025 Minor: N/A Page 4 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-95 Comments /

Reference:

OI-20 Rev: Major: 011 Minor: N/A Page 5 of 5

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-96 Examination Outline Cross-reference: 96 Revision: 1 Date: 11/30/20 Tier: 3 Group: N/A K/A Number: 2.2.25 Level of Difficulty: 2 SRO Importance Rating: 4.2 K/A

Description:

Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

What is the basis for the Reactor Coolant System (RCS) Pressure Safety Limit (SL)?

Maintaining RCS pressure below the SL ensures that RPV pressure will remain LE (1) of design pressure at the (2) .

A. (1) 110%

(2) reactor steam dome B. (1) 110%

(2) lowest elevation of the RCS C. (1) 125%

(2) reactor steam dome D. (1) 125%

(2) lowest elevation of the RCS Answer: B K/A Match:

Requires knowledge of the technical specification bases for safety limits.

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-96 SRO Only:

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-96 Explanation:

A. Incorrect. Plausible since the SL maintains pressure LE 110% of design pressure and the highest pressure in the RPV will be located at the lowest elevation. However, this is based on pressure at the lowest elevation of the RCS.

B. Correct. In accordance with the bases for Technical Specification 2.1.2, the limiting condition for RCS pressure is maintaining the RPV LE 110% of design pressure at the lowest elevation of the RCS.

C. Incorrect. Plausible since 125% of the design pressure is the limit for RCS components outside the RPV and RCS pressure is measured at the reactor steam dome. However, the stem asks for the RPV limit (110%) which is based on pressure at the lowest elevation of the RCS.

D. Incorrect. Plausible since the pressure limit is based on pressure at the lowest elevation of the RCS. Plausibility is enhanced since 125% of the design pressure is the limit for RCS components outside the RPV. However, the limit for the RPV is 110% of design pressure.

Technical Reference(s)

Technical Specification Bases Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 13427 - Describe the bases for the Reactor Steam Dome Pressure Safety Limit.

[TS Bases] (SRO-only).

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Examinee must know basis for RPV pressure safety limit.

10 CFR Part 55 Content: 55.43 2

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-96 Comments /

Reference:

Technical Specification Bases Rev: Major: 114 Minor: N/A

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-96

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Examination Outline Cross-reference: 97 Revision: 0 Date: 11/16/20 Tier: 3 Group: N/A K/A Number: 2.2.40 Level of Difficulty: 3 SRO Importance Rating: 4.7 K/A

Description:

Ability to apply Technical Specifications for a system.

Plant conditions are follows:

All Intermediate Range instruments are on Range 1.

Using the reference provided, what actions are required by Technical Specifications?

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 A. Suspend control rod withdrawals immediately.

B. Suspend CORE ALTERATIONS except control rod insertions immediately.

C. Fully insert all insertable control rods within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

D. Restore required SRMs to OPERABLE status within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Answer: D K/A Match:

Requires ability to apply Technical Specifications to SRMs.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Explanation:

A. Incorrect. Plausible if it is believed that more than one required SRM is inoperable. However, since TS table 3.3.1.2-1 requires only 3 SRM channels to be operable for the conditions given in the stem, only one of the inoperable SRMs is required and TS LCO 3.3.1.2, condition A applies.

B. Incorrect. Plausible if it is believed that the reactor is in Mode 5. However, with the Reactor Mode Switch in START/HOT-STBY, the reactor is in Mode 2 and LCO 3.3.1.2, condition A applies.

C. Incorrect. Plausible since this would be the correct action if the reactor were in Mode 3 or 4.

However, with the Reactor Mode Switch in START/HOT-STBY, the reactor is in Mode 2 and LCO 3.3.1.2, condition A applies.

D. Correct. In accordance with LCO 3.3.1.2, the reactor is in Mode 2 when the Reactor Mode Switch is in START/HOT-STBY. Table 3.3.1.2-1 states that 3 SRMs are required to be operable for the conditions given. Both SRM-B and and SRM-D are inoperable, which leaves two operable channels. Therefore LCO 3.3.1.2, condition A applies which requires that one of the two inoperable SRM channel be returned to operable within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Technical Reference(s)

Technical Specifications Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: LCO 3.3.1.2 actions and table 3.3.1.2-1 Learning Objective: 5944 - Referencing Columbia Generation Station Technical Specifications (section 3 only for initial license candidates) associated with the Source Range Monitoring System and a set of plant conditions, determine as applicable the LSSS, the LCO, the action statement, and the appropriate bases.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: None Question Cognitive Level:

Justification for Cognitive Level Requires candidate to synthesize information given in the question stem with a knowledge of reactor modes and an understanding of TS application for SRMs.

10 CFR Part 55 Content: 55.43 2 Page 4 of 11

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Comments /

Reference:

Technical Specifications Rev: Major: 254 Minor: N/A Page 5 of 11

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Page 6 of 11

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Page 7 of 11

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Proposed Reference Page 8 of 11

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-97 Page 9 of 11

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Examination Outline Cross-reference: 98 Revision: 0 Date: 8/26/20 Tier: 3 Group: N/A K/A Number: 2.3.6 Level of Difficulty: 2 SRO Importance Rating: 3.8 K/A

Description:

Ability to approve release permits CGS is in Mode 1.

Circulating Water (CW) Blowdown is required to be initiated.

Who is required to approve initiating CW Blowdown?

CW Blowdown initiation must be approved by the CRS/Shift Manager A. ONLY.

B. and Chemistry Manager.

C. and Operations Manager.

D. and Chemistry and Radiological Safety Manager.

Answer: A K/A Match:

Requires knowledge of requirements to approve a release permit.

SRO Only:

From NUREG 1021, ES-401:

Explanation:

A. Correct. In accordance with PPM 12.2.9, Circulating and Plant Service Water Halogenation Surveillance, step 8.5.6, the CRS/Shift Manager approves the CW Blowdown.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 B. Incorrect. Plausible since PPM 16.10.1, Radioactive Liquid Waste Discharge to the River, states that Chemistry Supervision must approve a liquid discharge if projected doses are greater than the limit. Additionally, PPM 12.2.9 requires informing Chemistry Management of certain conditions. However, CW Blowdown is not considered a radioactive release and Chemistry Management is not required to approve the blowdown.

C. Incorrect. Plausible since PPM 16.10.1 states that Liquid discharges to the river are to be avoided if at all possible, even if technically allowable. This suggests that release authorization from senior management is required. However, CW Blowdown is not considered a radioactive release and the Operations Manager is not required to approve the blowdown.

D. Incorrect. Plausible since PPM 16.10.1 states that the Chemistry and Radiological Safety Manager must approve a liquid discharge if the Effluent Concentration is > 3.0 Effluent Concentration Limit (ECL). However, CW Blowdown is not considered a radioactive release and the Chemistry and Radiological Safety Manager is not required to approve the blowdown.

Technical Reference(s)

PPM 12.2.9, Circulating and Plant Service Water Halogenation Surveillance Attached w/ Revision #

PPM 16.10.1, Radioactive Liquid Waste Discharge to the River See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: None Learning Objective: 11260 - Knowledge of the requirements for reviewing and approving release permits.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of individual required to authorize a CW Blowdown release permit.

10 CFR Part 55 Content: 55.43 4 Page 2 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Comments /

Reference:

PPM 12.2.9 Rev: Major: 044 Minor: 002 Page 3 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Page 4 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Page 5 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Comments /

Reference:

PPM 12.2.9 Rev: Major: 008 Minor: N/A Page 6 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-98 Page 7 of 7

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-99 Examination Outline Cross-reference: 99 Revision: 0 Date: 10/20/20 Tier: 3 Group: N/A K/A Number: 2.4.18 Level of Difficulty: 2 SRO Importance Rating: 4.0 K/A

Description:

Knowledge of the specific bases of the EOPs.

Which of the following is an assumption used to determine Cold Shutdown Boron Weight (CSBW)?

A. Reactor water is at 212°F.

B. All control rods are full out.

C. 50% power equilibrium Xenon in the core.

D. RWCU with filter demineralizers is in service.

Answer: B K/A Match:

Requires knowledge of the bases for determining CSBW which is used in PPM 5.1.2, RPV Control -

ATWS.

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-99 SRO Only:

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ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-99 Explanation:

A. Incorrect. Plausible since an assumption for reactor water temperature is used when determining CSBW. However, reactor water temperature is assumed to be at its most reactive temperature (68°F).

B. Correct. In accordance with PPM 5.0.10, Flowchart Training Manual, section 7.1, CSBW is determined assuming all control rods are full out.

C. Incorrect. Plausible since an assumption for Xenon in the core is used when determining CSBW.

However, it is assumed that no Xenon is in the core.

D. Incorrect. Plausible since an assumption for RWCU operation is used when determining CSBW.

However, filter demineralizers are assumed to be bypassed.

Technical Reference(s)

PPM 5.0.10, Flowchart Training Manual Attached w/ Revision #

Tech Ref 2 See Comments / Reference Tech Ref 3 Tech Ref 4 Proposed references to be provided during examination: N/A Learning Objective: 9822 - Using knowledge of the emergency operating procedure bases, discuss the need to predict plant response to recommended actions.

Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires knowledge of assumptions for determining CSBW.

10 CFR Part 55 Content: 55.43 5 Page 3 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-99 Comments /

Reference:

PPM 5.0.10 Rev: Major: 0232 Minor: N/A Page 4 of 4

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-100 Examination Outline Cross-reference: 100 Revision: 1 Date: 11/17/20 Tier: 3 Group: N/A K/A Number: 2.4.29 Level of Difficulty: 2 SRO Importance Rating: 4.4 K/A

Description:

Knowledge of the emergency plan.

CGS is in Mode 1.

An event occurs that requires entry into the emergency plan and evaluation of Emergency Action Levels (EALs).

The following conditions exist:

RPV Water Level: -165 inches, down slow. Level cannot be raised.

Drywell and Wetwell pressures are equal at 18 psig, up slow.

No drywell or wetwell spray in progress.

RHR-P-2B room temperature: 145°F, up slow.

Using the reference provided and considering the condition of Fission Product Barriers only, which EAL should be declared?

A. Unusual Event (UE)

B. Alert (A)

C. Site Area Emergency (SAE)

D. General Emergency (GE)

Answer: D K/A Match:

Requires knowledge of the emergency plan with respect to fission product barrier degradation.

SRO Only:

At CGS, EAL event classification is the exclusive responsibility of SRO-licensed individuals. See learning objective 6131 - With the procedures available for reference and plant conditions such that an emergency classification should be declared, correctly classify the event. (SRO only) [PPM 13.1.1]

From NUREG-1021, revision 11:

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-100 Explanation:

A. Incorrect. Plausible if it is believed that the loss of only one fission product barrier (FCB) is an unusual event. However, for the conditions given, a Site Area Emergency should be declared.

B. Incorrect. Plausible since a loss or potential loss of the fuel clad or RCS constitutes an Alert.

However, for the conditions given, a Site Area Emergency should be declared.

C. Incorrect. Plausible if it is not recognized that with drywell and wetwell pressures equal with no sprays is a PC pressure response not consistent with LOCA, and indicative of a failed drywell floor. The requirement to classify the event as a Site Area Emergency is the Loss or Potential Loss of any two barriers. Conditions given in the stem indicate a loss of two barriers and potential loss of the third barrier, which meets the criteria for GE.

D. Correct. The requirements for a General Emergency require the loss of two barriers and the loss or potential loss of the third barrier. Based on RPV level, there is a loss of the RCS barrier and potential loss of the PC barrier. Drywell and wetwell pressures equal with no sprays is a PC pressure response not consistent with LOCA, and indicative of a failed drywell floor.Therefor, there is a loss of two barriers and potential loss of the third barrier, which meets the criteria o of declaring a GE.

Technical Reference(s)

PPM 13.1.1, Classifying the Emergency Attached w/ Revision #

PPM 13.1.1A, Classifying the Emergency - Technical Basis See Comments / Reference PPM 5.3.1, Secondary Containment Control Tech Ref 4 Proposed references to be provided during examination: PPM 13.1.1 (complete), RB Area Temp data sheet.

Learning Objective: 6131 - With the procedures available for reference and plant conditions such that an emergency classification should be declared, correctly classify the event. (SRO only) [PPM 13.1.1]

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-100 Question Source: #: Bank #

Mod Bank #. (Note changes or attach parent)

Question History: Last NRC Exam: N/A Question Cognitive Level:

Justification for Cognitive Level Requires interpreting conditions given in the question stem against EAL requirements and a knowledge of fission barrier conditions that constitute different EAL classifications.

10 CFR Part 55 Content: 55.43 5 Comments /

Reference:

PPM 13.1.1 Rev: Major: 049 Minor: 001

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-100 Comments /

Reference:

PPM 13.1.1A Rev: Major: 034 Minor: 001

ES-401 CGS NRC 2021 SRO Written Exam Worksheet Form ES-401-5 Question: SRO-100 Comments /

Reference:

PPM 5.3.1 Rev: Major: 21 Minor: N/A

ILC-24 NRC Validation Exam Question SRO-79/100 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Prolonged loss of all offsite and all onsite AC power to emergency Loss of all offsite and all onsite AC power to emergency buses Loss of all but one AC power source to emergency buses Loss of all offsite AC power capability to emergency buses buses for 15 minutes or longer for 15 minutes or longer for 15 minutes or longer MG1.1 1 2 3 MS1.1 1 2 3 MA1.1 1 2 3 MU1.1 1 2 3 Loss of all offsite AND all onsite AC power capability to Loss of all offsite and all onsite AC power capability to AC power capability, Table 2, to emergency buses SM-7 Loss of all offsite AC power capability, Table 2, to emergency emergency buses SM-7 and SM-8 emergency buses SM-7 and SM-8 for GE 15 min. (Note 1) and SM-8 reduced to a single power source for GE 15 min. buses SM-7 and SM-8 for GE 15 min. (Note 1)

AND EITHER: (Note 1) 1 Restoration of emergency bus SM-7 or SM-8 in LT 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely (Note 1)

OR AND Any additional single power source failure will result in a loss of all AC power to SAFETY SYSTEMS Table 2 AC Power Sources Loss of Offsite Emergency RPV level cannot be restored and maintained GT -186 in. Startup Transformer TR-S AC Power Loss of all emergency AC and vital DC power sources for 15 Backup Transformer TR-B minutes or longer Backfeed 500 KV power through Main Transformers (if already MG1.2 1 2 3 aligned in modes 4, 5, def only)

Loss of all offsite AND all onsite AC power capability to Onsite emergency buses SM-7 and SM-8 for GE 15 min. (Note 1) DG1 AND DG2 2 Indicated voltage is LT 108 VDC on both 125 VDC buses Loss of all vital DC power for 15 minutes or longer Main Generator via TR-N1/N2 DP-S1-1 and DP-S1-2 for GE 15 min. (Note 1)

MS2.1 1 2 3 Loss of None None Indicated voltage is LT 108 VDC on both 125 VDC buses Vital DC DP-S1-1 and DP-S1-2 for GE 15 min. (Note 1)

Power UNPLANNED loss of Control Room indications for 15 minutes or UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress longer MA3.1 1 2 3 MU3.1 1 2 3 3 An UNPLANNED event results in the inability to monitor one or more Table 10 parameters from within the Control Room An UNPLANNED event results in the inability to monitor one or more Table 10 parameters from within the Control Loss of None None for GE 15 min. (Note 1) Room for GE 15 min. (Note 1)

Control AND Room Any Table 11 transient event in progress Indications Table 10 Safety System Parameters Reactor coolant activity greater than Technical Specification Reactor power allowable limits Table 5 Plant Structures Containing Safe Shutdown Systems or RPV level Components RPV pressure MU4.1 1 2 3 4

Primary containment pressure SJAE CONDSR OUTLET RAD HI-HI alarm (P602)

Vital portions of the Rad Waste/Control Building:

Wetwell level None None - 467' elevation vital island RCS Wetwell temperature Activity - 487' elevation cable spreading room MU4.2 1 2 3

- Main Control Room and vertical cable chase

- 525' elevation HVAC area Table 11 Transient Events Coolant activity GT 0.2 Ci/gm dose equivalent I-131 Reactor Building Vital portions of the Turbine Building Reactor scram M -

DEH pressure switches RPS switches on turbine throttle valves Runback GT 25% thermal reactor power RCS leakage for 15 minutes or longer System Electrical load rejection GT 25% full MU5.1 1 2 3

- Main steam line radiation monitors 5

Malfunct. electrical load (1) RCS unidentified or pressure boundary leakage

- Turbine Building ventilation radiation monitors ECCS injection GE 10 gpm for GE 15 min.

None - Main steam line piping up to MS-V-146 Noneand the first stop valves Thermal powerNone oscillations GT 10% OR RCS Leakage Standby Service Water Pump Houses (2) RCS identified leakage GT 25 gpm for GE 15 min.

Diesel Generator Building OR (3) Leakage from the RCS to a location outside containment GT 25 gpm for GE 15 min.

Inability to shut down the reactor causing a challenge to RPV Automatic or manual scram fails to shut down the reactor, and Automatic or manual scram fails to shut down the reactor water level or RCS heat removal subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor MS6.1 1 2 MA6.1 1 2 MU6.1 1 2 An automatic OR manual scram fails to shut down the An automatic OR manual scram fails to shut down the An automatic OR manual scram did not shut down the reactor reactor reactor 6 None AND All actions to shut down the reactor are not successful as AND Manual scram actions taken at the reactor control console AND A subsequent automatic scram OR manual scram action RPS indicated by reactor power GT 5% (mode switch in shutdown, manual push buttons or ARI) are taken at the reactor control console (mode switch in Failure AND EITHER: not successful in shutting down the reactor as indicated by shutdown, manual push buttons or ARI) is successful in RPV level cannot be restored and maintained reactor power GT 5% (Note 8) shutting down the reactor as indicated by reactor power LE Table 4 Communication Methods above -186 in. or cannot be determined 5% (APRM downscale) (Note 8)

OR System Onsite ORO NRC WW temperature and RPV pressure cannot be maintained below the HCTL Plant Public Address (PA) System X Loss of all onsite or offsite communications capabilities Plant Telephone System X X MU7.1 1 2 3 7 Plant Radio System Operations and Security Channels X

None None (1) Loss of all Table 4 onsite communication methods OR Loss of (2) Loss of all Table 4 ORO communication methods Comm. Offsite calling capability from the X X OR Control Room via direct telephone (3) Loss of all Table 4 NRC communication methods Long distance calling capability on X X None the commercial phone system Hazardous event affecting a SAFETY SYSTEM needed for the current operating mode MA8.1 1 2 3 The occurrence of any Table 8 hazardous event Table 8 Hazardous Events AND Event damage has caused indications of degraded 8 None Seismic event Internal or external FLOODING event performance on one train of a SAFETY SYSTEM needed for the current operating mode AND EITHER:

None Hazardous Event High winds Event damage has caused indications of degraded Affecting Tornado strike performance to a second train of a SAFETY SYSTEM Safety FIRE needed for the current operating mode Systems OR EXPLOSION Event damage has resulted in VISIBLE DAMAGE to a Volcanic ash fallout second train of a SAFETY SYSTEM needed for the Other events with similar hazard current operating mode characteristics as determined by the Shift (Notes 9, 10)

Manager F FG1.1 1 2 Loss of any two barriers 3 FS1.1 1 2 3 Loss or potential loss of any two barriers (Table F-1)

FA1.1 1 2 3 Any loss or any potential loss of EITHER Fuel Clad or RCS None Fission Product AND barrier (Table F-1)

Barrier Degradation Loss or potential loss of the third barrier (Table F-1)

Table F-1 Fission Product Barrier Threshold Matrix FC - Fuel Clad Barrier RCS - Reactor Coolant System Barrier PC - Containment Barrier Loss Potential Loss Loss Potential Loss Loss Potential Loss RPV level cannot be restored and RPV level cannot be restored and A SAG entry required maintained GT -161 in. maintained GT -161 in. None None SAG entry required RPV Water Level or cannot be determined. or cannot be determined.

UNISOLABLE break in any of the UNISOLABLE primary system leakage UNISOLABLE primary system leakage following:

that results in exceeding EITHER: that results in exceeding EITHER:

Main Steam Line RCIC Steam Line RB area temperature alarm level (PPM RB area maximum safe operating B None None RWCU 5.3.1 Table 23) temperature (PPM 5.3.1 Table 23) None RCS Leak Rate Feedwater OR OR OR RB area radiation alarm level (PPM RB area maximum safe operating Emergency RPV Depressurization is 5.3.1 Table 24) radiation (PPM 5.3.1 Table 24) required PC pressure GT 45 psig OR UNPLANNED rapid drop in PC pressure following PC pressure rise Explosive mixture exists inside PC C None None PC pressure GT 1.68 psig due to RCS None OR (H2 GE 6% and O2 GE 5%)

PC Conditions leakage OR PC pressure response not consistent with LOCA conditions WW temperature and RPV pressure cannot be maintained below the HCTL Containment Radiation Monitor CMS-RIS-27E or CMS-RIS-27F D reading GT 3,600 R/hr Containment Radiation Monitor Containment Radiation Monitor None CMS-RIS-27E or CMS-RIS-27F None None CMS-RIS-27E or CMS-RIS-27F PC Rad / OR reading GT 70 R/hr reading GT 14,000 R/hr RCS Activity Primary coolant activity GT 300 µCi/gm Dose Equivalent I-131 UNISOLABLE direct downstream E pathway to the environment exists after PC Integrity or None PC isolation signal None None None None Bypass OR Intentional PC venting per EOPs F Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion of the Any condition in the opinion of the Emergency Emergency Director that indicates Emergency Director that indicates loss Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates loss Emergency Director that indicates loss Director potential loss of the Fuel Clad barrier of the RCS barrier potential loss of the RCS barrier potential loss of the Containment of the fuel clad barrier of the Containment barrier Judgment barrier 13.1.1 Rev. 49 MR 1 CLASSIFYING THE EMERGENCY Modes: 1 2 3 HOT CONDITIONS 1/16/2019 Power Operations Startup Hot Shutdown (RCS GT 200°F) 170028

ILC-24 NRC Validation Exam Question SRO-79/100 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Loss of RPV inventory affecting fuel clad integrity with Loss of RPV inventory affecting core decay heat removal Significant loss of RPV inventory Unplanned loss of RPV inventory containment challenged capability CG1.1 4 5 CS1.1 4 5 CA1.1 4 5 CU1.1 4 5 RPV level LT -161 in. for GE 30 min. (Note 1) (1) CONTAINMENT CLOSURE not established (1) Loss of RPV inventory as indicated by RPV level (1) UNPLANNED loss of reactor coolant results in RPV level AND AND LT -50 in. less than a required lower limit for GE 15 min. (Note 1)

OR OR Any of the following indications of containment challenge: RPV level LT -129 in.

(2) RPV level cannot be monitored for GE 15 min. (Note 1) (2) RPV level cannot be monitored CONTAINMENT CLOSURE not established (Note 6) OR AND AND Explosive mixture inside PC (2) CONTAINMENT CLOSURE established UNPLANNED increase in any Table 1 sump or pool levels (H2 GE 6% and O2 GE 5%) UNPLANNED increase in any Table 1 sump or pool AND due to a loss of RPV inventory UNPLANNED rise in PC pressure levels due to a loss of RPV inventory RPV level LT -161 in.

RB area radiation GT any Maximum Safe Operating level (PPM 5.3.1 Table 24)

CS1.2 4 5 CG1.2 4 5 1

RPV level cannot be monitored for GE 30 min. (Note 1)

AND AND RPV Core uncovery is indicated by any of the following: Table 1 Sumps/Pool Level Core uncovery is indicated by any of the following:

UNPLANNED wetwell level rise GT 2 inches UNPLANNED wetwell level rise GT 2 inches (PPM 5.2.1 entry condition) Any valid Hi-Hi level alarm on R-1 Table 2 AC Power Sources (PPM 5.2.1 entry condition) VALID indication of RB room flooding as identified by through R-5 sumps VALID indication of RB room flooding as identified by high level alarms (PPM 5.3.1 Table 25) EDR GE 25 GPM high level alarms (PPM 5.3.1 Table 25) Offsite Observation of UNISOLABLE RCS leakage outside FDR GE 10 GPM Observation of UNISOLABLE RCS leakage outside Startup Transformer TR-S primary containment of sufficient magnitude to indicate primary containment of sufficient magnitude to indicate core uncovery Wetwell level rise Backup Transformer TR-B core uncovery Observation of UNISOLABLE RCS Backfeed 500 KV power through Main AND leakage Transformers (if already aligned in Any of the following indications of containment challenge: modes 4, 5, def only)

CONTAINMENT CLOSURE not established (Note 6)

Explosive mixture inside PC Onsite (H2 GE 6% and O2 GE 5%) DG1 UNPLANNED rise in PC pressure DG2 RB area radiation GT any Maximum Safe Operating Main Generator via TR-N1/N2 level (PPM 5.3.1 Table 24)

Loss of all offsite and all onsite AC power to emergency buses Loss of all but one AC power source to emergency buses for 15 for 15 minutes or longer minutes or longer 2 None None CA2.1 4 5 Loss of all offsite and all onsite AC power capability to DEF CU2.1 4 5 AC power capability, Table 2, to emergency buses SM-7 DEF Loss of emergency buses SM-7 and SM-8 for GE 15 min. (Note 1) and SM-8 reduced to a single power source for GE 15 min.

C Emergency AC Power (Note 1)

AND Any additional single power source failure will result in a loss Cold SD/

of all AC power to SAFETY SYSTEMS Refuel System Malfunct. Inability to maintain plant in cold shutdown UNPLANNED increase in RCS temperature Table 7 RCS Reheat Duration Thresholds 4 5 4 5 3

CA3.1 CU3.1 UNPLANNED increase in RCS temperature to GT 200°F UNPLANNED increase in RCS temperature to GT 200°F

If an RCS heat removal system is in operation within this None for GT Table 7 duration (Note 1)

RCS time frame and RCS temperature is being reduced the EAL is not applicable OR Temp.

UNPLANNED RPV pressure increase GT 10 psig CU3.2 4 5 Containment Heat-up RCS Status Closure Status Duration Loss of all RCS temperature and RPV water level indication for GE 15 min. (Note 1)

Intact N/A 60 min.

4 None Not intact established 20 min.

None CU4.1 Loss of vital DC power for 15 minutes or longer 4 5 Loss of not established 0 min.

Vital DC Indicated voltage LT 108 VDC on required 125 VDC buses Power DP-S1-1 and DP-S1-2 for GE 15 min. (Note 1)

Loss of all onsite or offsite communications capabilities Table 4 Communication Methods CU5.1 4 5 DEF 5 System None Onsite ORO NRC None None Loss of all Table 4 onsite communication methods OR Loss of Comm. Plant Public Address (PA) System X Loss of all Table 4 ORO communication methods OR Plant Telephone System X X Loss of all Table 4 NRC communication methods Hazardous event affecting a SAFETY SYSTEM needed for the Plant Radio System Operations and X current operating mode Security Channels CA6.1 4 5 Offsite calling capability from the X X The occurrence of any Table 8 hazardous event Table 8 Hazardous Events Control Room via direct telephone AND 6

None None Event damage has caused indications of degraded Long distance calling capability on X X Seismic event performance on one train of a SAFETY SYSTEM needed for Hazardous the commercial phone system Internal or external FLOODING event the current operating mode Events High winds AND EITHER:

Affecting Event damage has caused indications of degraded Safety Tornado strike performance to a second train of a SAFETY SYSTEM Systems needed for the current operating mode FIRE EXPLOSION OR Event damage has resulted in VISIBLE DAMAGE to a Volcanic ash fallout second train of a SAFETY SYSTEM needed for the Other events with similar hazard current operating mode characteristics as determined by the Shift (Notes 9, 10)

Manager 13.1.1 Rev. 49 MR 1 CLASSIFYING THE EMERGENCY Modes: 4 5 DEF COLD CONDITIONS 1/16/2019 Cold Shutdown Refueling Defueled (RCS 200°F) 170028

ILC-24 NRC Validation Exam Question SRO-79/100 GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous or liquid radioactivity resulting in offsite Release of gaseous or liquid radioactivity greater than 2 times than 1,000 mrem TEDE or 5,000 mrem thyroid CDE than 100 mrem TEDE or 500 mrem thyroid CDE dose greater than 10 mrem TEDE or 50 mrem thyroid CDE the ODCM limits for 60 minutes or longer RG1.1 1 2 3 4 5 DEF RS1.1 1 2 3 4 5 DEF RA1.1 1 2 3 4 5 DEF RU1.1 1 2 3 4 5 DEF (1) Reading on any Table 3 effluent radiation monitor (1) Reading on any Table 3 effluent radiation monitor (1) Reading on any Table 3 effluent radiation monitor (1) Reading on any Table 3 effluent radiation monitor GT column "GENERAL" for GE 15 min. GT column "SAE" for GE 15 min. GT column "ALERT" for GE 15 min. GT column "UE" for GE 60 min.

OR OR OR OR (2) Dose assessment using actual meteorology indicates (2) Dose assessment using actual meteorology (2) Dose assessment using actual meteorology indicates (2) Sample analyses for a gaseous or liquid release doses GT 1,000 mrem TEDE or GT 5000 mrem thyroid indicates doses GT 100 mrem TEDE or GT 500 doses GT 10 mrem TEDE or GT 50 mrem thyroid CDE indicates a concentration or release rate 2 x ODCM CDE at or beyond the SITE BOUNDARY mrem thyroid CDE at or beyond the SITE at or beyond the SITE BOUNDARY limits for GE 60 min.

(Notes 1, 2, 3, 4) BOUNDARY (Notes 1, 2, 3, 4) (Notes 1, 2, 3) 1 (Notes 1, 2, 3, 4)

RA1.2 1 2 3 4 5 DEF RG1.2 1 2 3 4 5 DEF RS1.2 1 2 3 4 5 DEF Analysis of a liquid effluent sample indicates a concentration Rad Field survey results indicate EITHER of the following at or Field survey results indicate EITHER of the following at or or release rate that would result in doses GT 10 mrem TEDE Effluent beyond the SITE BOUNDARY: beyond the SITE BOUNDARY: or GT 50 mrem thyroid CDE at or beyond the SITE Closed window dose rates GT 100 mR/hr expected BOUNDARY for 60 min. of exposure (Notes 1, 2)

Closed window dose rates GT 1,000 mR/hr expected to continue for GE 60 min. to continue for GE 60 min.

Analyses of field survey samples indicate thyroid RA1.3 1 2 3 4 5 DEF Analyses of field survey samples indicate thyroid CDE GT 5,000 mrem for 60 min. of inhalation. CDE GT 500 mrem for 60 min. of inhalation.

Field survey results indicate EITHER of the following at or (Notes 1, 2) (Notes 1, 2) beyond the SITE BOUNDARY:

Closed window dose rates GT 10 mR/hr expected to continue for GE 60 min.

Analyses of field survey samples indicate thyroid CDE GT 50 mrem for 60 min. of inhalation.

R (Notes 1, 2)

Abnormal Spent fuel pool level cannot be restored to at least the top of the Spent fuel pool level at the top of the fuel racks Significant lowering of water level above, or damage to, Unplanned loss of water level above irradiated fuel Rad fuel racks for 60 minutes or longer irradiated fuel Levels RG2.1 1 2 3 4 5 DEF RS2.1 1 2 3 4 5 DEF RA2.1 1 2 3 4 5 DEF RU2.1 1 2 3 4 5 DEF

/

Rad Spent fuel pool level cannot be restored to at least 0.5 ft Lowering of spent fuel pool level to 0.5 ft Uncovery of irradiated fuel in the REFUELING PATHWAY UNPLANNED water level drop in the REFUELING PATHWAY Effluent for GE 60 min. (Note 1) as indicated by EITHER of the following:

RA2.2 1 2 3 4 5 DEF 2

SFP level LE 22.3 ft Damage to irradiated fuel resulting in a release of SFP low level alarm Table 3 Effluent Monitor Classification Thresholds radioactivity AND Irradiated AND UNPLANNED rise in area radiation levels as indicated by any Fuel Event Release Point Monitor General SAE Alert UE High alarm on any of the following radiation monitors: of the following radiation monitors:

ARM-RIS-1 Reactor Building Fuel Pool Area ARM-RIS-1 Reactor Building Fuel Pool Area PRM-RE-11 3.05E-03 µCi/cc ARM-RIS-2 Reactor Building Fuel Pool Area ARM-RIS-2 Reactor Building Fuel Pool Area Reactor Building Exhaust PRM-RE-12 ---- ---- 2.82E+1 µCi/cc ---- ARM-RIS-34 Reactor Building Elevation 606 ARM-RIS-34 Reactor Building Elevation 606 Gaseous PRM-RE-13 7.50E+02 µCi/cc 7.50E+1 µCi/cc ---- ---- REA-RIS-609A-D Rx Bldg Vent Turbine Building Exhaust TEA-RIS-13 8.35E-02 µCi/cc 8.35E-03 µCi/cc 8.35E-04 µCi/cc 4.22E-05 µCi/cc RA2.3 1 2 3 4 5 DEF Radwaste Building Exhaust WEA-RIS-14 3.45E-01 µCi/cc 3.45E-02 µCi/cc 3.45E-03 µCi/cc 3.98E-04 µCi/cc Lowering of spent fuel pool level to 10 ft Radwaste Effluent FDR-RIS-606 ---- ---- ---- 2 X HI-HI alarm Radiation levels that IMPEDE access to equipment necessary for normal plant operations, cooldown or shutdown 3 Liquid TSW Effluent TSW-RIS-5 ---- ---- ---- 3.00E-05 µCi/cc RA3.1 1 2 3 4 5 DEF Service Water Process A SW-RIS-604 ---- ---- ---- 1.00E+02 cps Area SW-RIS-605 1.00E+02 cps (1) Dose rates GT 15 mR/hr in Control Room Service Water Process B ---- ---- ----

Radiation (ARM-RIS-19) or CAS (by survey)

Levels OR (2) An UNPLANNED event results in radiation levels that prohibit or IMPEDE access to any Table 9 Table 9 Safe Operation & Shutdown Rooms/Areas rooms or areas (Note 5)

Damage to a loaded cask CONFINEMENT BOUNDARY Room/Area Modes Applicability EU1.1 Storage Operations RW 467 Radwaste Control Room (RHR flush to RW tanks) 3 1

Damage to a loaded canister (MPC) CONFINEMENT E Confinement RW 467 Vital Island (RHR-V-9 disconnect)

RB 422 B RHR Pump Rm (local pump temperatures) 3 3

BOUNDARY as indicated by measured dose rates on a loaded overpack GT EITHER:

ISFSI None RB 454 B RHR Pump Rm (operate RHR-V-85B) None 3 None 20 mrem/hr (gamma + neutron) on the top of the Boundary overpack 100 mrem/hr (gamma + neutron) on the side of the overpack, excluding inlet and outlet ducts HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION within the OWNER CONTROLLED AREA or Confirmed SECURITY CONDITION or threat airborne attack threat within 30 minutes HS1.1 1 2 3 4 5 DEF HA1.1 1 2 3 4 5 DEF HU1.1 1 2 3 4 5 DEF A HOSTILE ACTION is occurring or has occurred within (1) A HOSTILE ACTION is occurring or has occurred (1) A SECURITY CONDITION that does not involve a the PROTECTED AREA as reported by the Security within the OWNER CONTROLLED AREA as reported HOSTILE ACTION as reported by the Security 1 None Sergeant or Security Lieutenant by the Security Sergeant or Security Lieutenant OR Sergeant or Security Lieutenant OR Security (2) A validated notification from NRC of an aircraft attack (2) Notification of a credible security threat directed at the threat within 30 min. of the site site OR (3) A validated notification from the NRC providing information of an aircraft threat Seismic event GT OBE levels 2 None None See CA6.1/MA8.1 for potential for upgrade to an AlertNone based on degraded HU2.1 1 2 3 4 5 DEF Seismic safety system performance or damage Seismic event GT Operating Basis Earthquake (OBE) as Event indicated by H13.P851.S1.5-1 (OPERATING BASIS EARTHQUAKE EXCEEDED) activated Notes Hazardous event HU3.1 1 2 3 4 5 DEF 1 The Emergency Director should declare the event promptly upon determining that time limit has been (1) A tornado strike within the PROTECTED AREA exceeded, or will likely be exceeded OR 2 If an ongoing release is detected and the release start (2) Volcanic ash fallout requiring plant shutdown time is unknown, assume that the release duration has exceeded the specified time limit HU3.2 1 2 3 4 5 DEF 3 If the effluent flow past an effluent monitor is known to 3 have stopped, indicating that the release path is isolated, the effluent monitor reading is no longer VALID for None See CA6.1/MA8.1 for potential for upgrade to an AlertNone based on degraded safety system performance or damage Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical isolation of a SAFETY Natural or classification purposes SYSTEM component needed for the current operating mode Tech. 4 The pre-calculated effluent monitor values presented in Hazard EALs RA1.1, RS1.1 and RG1.1 should be used for HU3.3 1 2 3 4 5 DEF emergency classification assessments until the results (1) Movement of personnel within the PROTECTED AREA is from a dose assessment using actual meteorology are IMPEDED due to an offsite event involving hazardous available materials (e.g., an offsite chemical spill, 618-11 event or 5 If the equipment in the listed room or area was already toxic gas release) inoperable or out-of-service before the event occurred, OR then no emergency classification is warranted (2) A hazardous event that results in on-site conditions 6 If CONTAINMENT CLOSURE is re-established prior to sufficient to prohibit the plant staff from accessing the site exceeding the 30-minute time limit, declaration of a via personal vehicles (Note 7)

General Emergency is not required 7 This EAL does not apply to routine traffic impediments FIRE potentially degrading the level of safety of the plant such as fog, snow, ice, or vehicle breakdowns or Table 5 accidents Plant Structures Containing Safe Shutdown Systems or Components HU4.1 1 2 3 4 5 DEF 8 A manual scram action is any operator action, or set of actions, which causes the control rods to be rapidly A FIRE is not extinguished within 15 min. of any of the inserted into the core, and does not include manually Vital portions of the Rad Waste/Control Building: following FIRE detection indications (Note 1):

driving in control rods or implementation of boron injection Report from the field (i.e., visual observation)

- 467' elevation vital island Receipt of multiple (more than 1) fire alarms or strategies H

- 487' elevation cable spreading room indications 9 If the affected SAFETY SYSTEM train was already Field verification of a single fire alarm

- Main Control Room and vertical cable chase inoperable or out of service before the hazardous event AND occurred, then emergency classification is not warranted - 525' elevation HVAC area The FIRE is located within any Table 5 area Hazards 10 If the hazardous event only resulted in VISIBLE Reactor Building DAMAGE, with no indications of degraded performance Vital portions of the Turbine Building HU4.2 1 2 3 4 5 DEF 4 to at least one train of a SAFETY SYSTEM, then this emergency classificationNoneis not warranted None -

DEH pressure switches RPS switches on turbine throttle valves Receipt of a single fire alarm (i.e., no other indications of a FIRE)

Fire AND

- Main steam line radiation monitors The fire alarm is indicating a FIRE within any Table 5 area

- Turbine Building ventilation radiation monitors AND The existence of a FIRE is not verified within 30 min. of alarm

- Main steam line piping up to MS-V-146 and the first stop valves receipt (Note 1)

Standby Service Water Pump Houses Diesel Generator Building HU4.3 1 2 3 4 5 DEF (1) A FIRE within the ISFSI or plant PROTECTED AREA not extinguished within 60 min. of the initial report, alarm or indication (Note 1)

See CA6.1/MA8.1 for potential for OR upgrade to an Alert based on degraded (2) A FIRE within the ISFSI or plant PROTECTED AREA that safety system performance or damage requires firefighting support by an offsite fire response Table 9 Safe Operation & Shutdown Rooms/Areas agency to extinguish Room/Area Modes Applicability Gaseous release IMPEDING access to equipment necessary for normal plant operations, cooldown or shutdown 5

RW 467 Radwaste Control Room (RHR flush to RW tanks) 3 RW 467 Vital Island (RHR-V-9 disconnect) 3 HA5.1 1 2 3 4 5 DEF None None Hazardous RB 422 B RHR Pump Rm (local pump temperatures) 3 Release of a toxic, corrosive, asphyxiant or flammable gas Gases into any Table 9 rooms or areas RB 454 B RHR Pump Rm (operate RHR-V-85B) 3 AND Entry into the room or area is prohibited or IMPEDED (Note 5)

Inability to control a key safety function from outside the Control Control Room evacuation resulting in transfer of plant control to Room alternate locations HS6.1 1 2 3 4 5 HA6.1 1 2 3 4 5 DEF 6 None An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel or Alternate Remote Shutdown Panel An event has resulted in plant control being transferred from the Control Room to the Remote Shutdown Panel or Alternate Remote Shutdown Panel None Control AND Room Control of any of the following key safety functions is not Evacuation reestablished within 15 min. (Note 1):

Reactivity (Modes 1 and 2 only)

RPV water level RCS heat removal Other conditions existing which in the judgment of the Emergency Other conditions existing which in the judgment of the Emergency Other conditions existing which in the judgment of the Other conditions existing which in the judgment of the Director warrant declaration of General Emergency Director warrant declaration of Site Area Emergency Emergency Director warrant declaration of an Alert Emergency Director warrant declaration of a UE HG7.1 1 2 3 4 5 DEF HS7.1 1 2 3 4 5 DEF HA7.1 1 2 3 4 5 DEF HU7.1 1 2 3 4 5 DEF Other conditions exist which, in the judgment of the Other conditions exist which, in the judgment of the Other conditions exist which, in the judgment of the Other conditions exist which, in the judgment of the Emergency Director, indicate that events are in progress or Emergency Director, indicate that events are in progress or Emergency Director, indicate that events are in progress or Emergency Director, indicate that events are in progress or 7 have occurred which involve actual or IMMINENT have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the substantial core degradation or melting with potential for plant functions needed for protection of the public or substantial degradation of the level of safety of the plant or level of safety of the plant or indicate a security threat to loss of containment integrity or HOSTILE ACTION that HOSTILE ACTION that results in intentional damage or a security event that involves probable life threatening risk facility protection has been initiated. No releases of Judgment results in an actual loss of physical control of the facility. malicious acts, (1) toward site personnel or equipment that to site personnel or damage to site equipment because of radioactive material requiring offsite response or monitoring Releases can be reasonably expected to exceed EPA could lead to the likely failure of or, (2) that prevent effective HOSTILE ACTION. Any releases are expected to be limited are expected unless further degradation of SAFETY Protective Action Guideline exposure levels offsite for more access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline SYSTEMS occurs.

than the immediate site area. Any releases are not expected to result in exposure levels exposure levels.

which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.

13.1.1 Rev. 49 MR 1 CLASSIFYING THE EMERGENCY Modes: 1 2 3 4 5 DEF 1/16/2019 Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled ALL CONDITIONS 170028

ILC-24 NRC Validation Exam Question: SRO-86 ECCS - Operating 3.5.1 3.5 EMERGENCY CORE COOLING SYSTEMS (ECCS), RPV WATER INVENTORY CONTROL, AND REACTOR CORE ISOLATION COOLING (RCIC) SYSTEM 3.5.1 ECCS - Operating LCO 3.5.1 Each ECCS injection/spray subsystem and the Automatic Depressurization System (ADS) function of six safety/relief valves shall be OPERABLE.

APPLICABILITY: MODE 1, MODES 2 and 3, except ADS valves are not required to be OPERABLE with reactor steam dome pressure 150 psig.

ACTIONS

NOTE-----------------------------------------------------------

LCO 3.0.4.b is not applicable to High Pressure Core Spray (HPCS).

CONDITION REQUIRED ACTION COMPLETION TIME A. One low pressure ECCS A.1 Restore low pressure 7 days(1) injection/spray ECCS injection/spray subsystem inoperable. subsystem to OPERABLE status.

B High Pressure Core B.1 Verify by administrative Immediately Spray (HPCS) System means RCIC System is inoperable. OPERABLE when RCIC System is required to be OPERABLE.

AND B.2 Restore HPCS System to 14 days OPERABLE status.

(1) The Completion Time that one train of RHR (RHR-A) can be inoperable as specified by Required Action A.1 may be extended beyond the 7 day completion time up to 7 days to support restoration of RHR-A following pump and motor replacement. This footnote will expire at 23:59 PST February 28, 2019.

Columbia Generating Station 3.5.1-1 Amendment No. 187 225 230 245 251 253 1 of 15

ILC-24 NRC Validation Exam Question: SRO-86 ECCS - Operating 3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. Two ECCS injection C.1 Restore ECCS 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months subsystems inoperable. injection/spray subsystem to OPERABLE status.

OR One ECCS injection and one ECCS spray subsystem inoperable.

D. Required Action and D.1 --------------NOTE---------------

associated Completion LCO 3.0.4.a is not Time of Condition A, B, applicable when entering or C not met. MODE 3.

Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months E. One required ADS valve E.1 Restore ADS valve to 14 days inoperable. OPERABLE status.

F. One required ADS valve F.1 Restore ADS valve to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months inoperable. OPERABLE status.

AND OR One low pressure ECCS F.2 Restore low pressure 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months injection/spray ECCS injection/spray subsystem inoperable. subsystem to OPERABLE status.

Columbia Generating Station 3.5.1-2 Amendment No. 149,169,225,236 2 of 15

ILC-24 NRC Validation Exam Question: SRO-86 ECCS - Operating 3.5.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME G. Required Action and G.1 --------------NOTE--------------

associated Completion LCO 3.0.4.a is not Time of Condition E or F applicable when entering not met. MODE 3.

OR Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Two or more required ADS valves inoperable.

H. HPCS and Low H.1 Enter LCO 3.0.3. Immediately Pressure Core Spray (LPCS) Systems inoperable.

OR Three or more ECCS injection/spray subsystems inoperable.

OR HPCS System and one or more required ADS valves inoperable.

OR Two or more ECCS injection/spray subsystems and one or more required ADS valves inoperable.

Columbia Generating Station 3.5.1-3 Amendment No. 149,169,225,236 3 of 15

ILC-24 NRC Validation Exam Question: SRO-86 ECCS - Operating 3.5.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.5.1.1 Verify, for each ECCS injection/spray subsystem, In accordance locations susceptible to gas accumulation are with the sufficiently filled with water. Surveillance Frequency Control Program SR 3.5.1.2 ------------------------------NOTE---------------------------

Not required to be met for system vent flow paths opened under administrative controls.

Verify each ECCS injection/spray subsystem In accordance manual, power operated, and automatic valve in the with the flow path, that is not locked, sealed, or otherwise Surveillance secured in position, is in the correct position. Frequency Control Program SR 3.5.1.3 Verify ADS accumulator backup compressed gas In accordance system average pressure in the required bottles is with the 2200 psig. Surveillance Frequency Control Program SR 3.5.1.4 Verify each ECCS pump develops the specified flow In accordance rate with the specified differential pressure between with the reactor and suction source. INSERVICE TESTING DIFFERENTIAL PROGRAM PRESSURE BETWEEN REACTOR AND SYSTEM FLOW RATE SUCTION SOURCE LPCS 6200 gpm 128 psid LPCI 7200 gpm 26 psid HPCS 6350 gpm 200 psid Columbia Generating Station 3.5.1-4 Amendment No. 169,205,225,229,236 238 243 246 249 4 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Number: OSP-LPCS/IST-Q702 Use Category: CONTINUOUS Major Rev: 044 Minor Rev: 001

Title:

LPCS System Operability Test Page: 6 of 36 4.0 PRECAUTIONS AND LIMITATIONS 4.1 Prior to starting LPCS-P-1, assure discharge piping is pressurized by verifying H13-P601.A3-5.3, LPCS PUMP DISH PRESS HIGH/LOW alarm is not lit.

4.2 Maintain piping systems pressurized during testing by fully closing LPCS-V-12 prior to stopping pump.

4.3 Do not operate LPCS-P-1 without minimum flow requirements satisfied (770 gpm). {C-9448}

4.4 Watch for pump cavitation during the opening stroke of LPCS-V-12; if fluctuations in pump amperes, system flow or pressure occur, apply a close signal to the valve until fluctuations cease.

4.5 When it is desired to limit the testing to obtain specific test data, exercise care to ensure that the portion(s) of the procedure being performed provides the test data in the same manner as if the entire procedure were performed.

4.6 Alert Range 4.6.1 Pumps Measured test parameters beyond the Alert Value shall have the test frequency increased to once each 46 days until the cause of the deviation is determined and the condition is corrected, at which time the original test frequency may be resumed. Any abnormality or erratic action shall be evaluated per the Corrective Action Program. {R-9653}

4.6.2 Valves Valves with measured stroke times beyond the Alert Value shall be immediately retested or declared inoperable. {R-9653}

If the valve is retested and the second set of data is also beyond the Alert Value, the data shall be analyzed within 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months to verify that the new stroke time represents acceptable valve operation, or the valve shall be declared inoperable. {R-9653}

If the second set of data is within the Acceptance Range, the cause of the initial deviation shall be analyzed. {R-9653}

Any abnormality or erratic actions shall be evaluated per the Corrective Action program. Reference the Condition Report on the procedure cover sheet. {R-9653}

5 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Number: OSP-LPCS/IST-Q702 Use Category: CONTINUOUS Major Rev: 044 Minor Rev: 001

Title:

LPCS System Operability Test Page: 7 of 36 4.7 Action Range If a valve fails to open or close, or the measured test parameters are beyond the Action Value, declare the pump or valve inoperable. Any abnormality or erratic action shall be evaluated per the Corrective Action program. {R-9653}

4.8 Stroke Timing When stroke timing a valve open or closed, start the stop watch when the control switch is turned to the open or closed position and stopped when the valve indicates full open or closed.

When stroke timing a valve open that has both lights illuminated during its intermediate position, ensure the green light is extinguished prior to stopping the stopwatch and when stroke timing the valve closed, ensure the red light is extinguished prior to stopping the stopwatch.

4.9 Valve motors may be damaged by stroking too frequently or applying repeated start signals to a stalled valve. When possible, investigate malfunctions locally prior to repeating a valve motor energization.

4.10 Valves are to be tested in the As Found condition unless specifically directed otherwise.

Clearly specify the reason for not stroke timing any valve in the As Found Condition in the Comments Section.

4.11 Notify the Shift Manager before performing any step or section of this surveillance which affects the Control Room Operator's ability to control the LPCS system.

4.12 Note all discrepancies encountered during the performance of this surveillance on the cover sheet and report to the Shift Manager.

4.13 Unless specified in this surveillance makes no internal or external adjustments to plant equipment.

4.14 Ensure all trip, alarm, and computer points are clear after a channel is returned to service.

4.15 The unavailability of LPCS when it is required to be operable is required to be recorded.

The time when LPCS is declared unavailable should be minimized. (M-Rule) 4.16 ASME Code requires performance of a Comprehensive Pump Test (CPT) every two years. The CPT will use more accurate pressure instrumentation and has more restrictive acceptance criteria.

4.17 IST acceptance criteria is more limiting than the Technical Specification SR 3.5.1.4 acceptance criteria for LPCS-P-1.

4.18 Only qualified I&C Technicians should attempt to re-zero instruments used in this surveillance.

6 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Number: OSP-LPCS/IST-Q702 Use Category: CONTINUOUS Major Rev: 044 Minor Rev: 001

Title:

LPCS System Operability Test Page: 8 of 36 4.19 Infrared temperature detectors, like the Fluke laser unit or equivalent, are the preferred device for verification of adequate seal flow. This allows a reduction in dose and gives a larger margin of safety due to distance from rotating equipment.

4.20 When using infrared temperature detectors the target area should not have a reflective surface and the laser gun should be held within 2 feet of the target to ensure accurate readings.

4.21 Per the FSAR, minimum suppression pool temperature should be GT 50 °F per LDCN-14-040 to Table 6.2-19. It is suggested to maintain suppression pool temperatures between 55 - 90°F.

7 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Number: OSP-LPCS/IST-Q702 Use Category: CONTINUOUS Major Rev: 044 Minor Rev: 001

Title:

LPCS System Operability Test Page: 26 of 36 VALVE STROKE DATA SHEET

OPENING TIME IN SECONDS # CLOSING TIME IN SECONDS VALVE ID Ref. Alert Lo Alert Hi Action Hi Ref. Alert Lo Alert Hi Action Hi Measured Value Measured Value Value (+1)(+2) (+1)(+2) (+1)(+2) Value (+1)(+2) (+1)(+2) (+1)(+2)

LPCS-V-1 118 100 136 153 114 97 131 148 H

LPCS-FCV-11 12 10 14 16 10 8 13 H 15 LPCS-V-12 18 N/A N/A N/A 18 15 21 23 (+3)

H LPCS-V-3 NOT NOT N/A N/A N/A N/A N/A N/A OPEN CLOSED

(+1) For measured values beyond the Alert Value or Action Value refer to Precaution and Limitations 4.6 or 4.7, respectively.

(+2) When comparing measured values to Alert and Action limits round all measured Stroke Times to the nearest second. Use standard rounding techniques e.g., 17.49 rounds to 17 and 17.5 rounds to 18 seconds.

(+3) Use listed closing stroke time as limiting even though a higher limit is specified in Technical Specification.

H Motor operated valve.

END Attachment 9.1, Valve Stroke Data Sheet 8 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Number: OSP-LPCS/IST-Q702 Use Category: CONTINUOUS Major Rev: 044 Minor Rev: 001

Title:

LPCS System Operability Test Page: 36 of 36 LPCS-P-1 Acceptance Criteria (CPT)

LPCS-P-1 ACCEPTANCE CRITERIA (Two Year CPT) 340 1

338 336 ACTION RANGE 334 332 330 2

328 326 324 322 320 Differential Pressure in PSID 318 316 REFERENCE CURVE 314 312 310 308 306 304 5

302 300 298 4

296 ALERT RANGE 6 294 292 290 ACTION RANGE 3 288 286 6,375 6,400 6,425 6,450 6,475 6,500 6,525 6,550 6,575 6,600 6,625 Indicated Pump Flowrate in GPM ALERT RANGE = Area Inside 3-4-5-6 ACTION RANGE = Area Outside 1-2-3-4 END Attachment 9.8, LPCS-P-1 ACCEPTANCE CRITERIA (Two Year CPT) 9 of 15

ILC-24 NRC Validation Exam Question: SRO-86 Table 1.6.1.3-1 (page 3 of 19)

Primary Containment Isolation Valves MAXIMUM PEN VALVE ISOLATION TIME VALVE VALVE TYPE NUMBER NUMBER (Seconds) GROUP(a) CODE NOTES 24 FDR-V-3 15 4 AIV 24 FDR-V-4 15 4 AIV 101 FPC-V-149 35 4 AIV 101 FPC-V-156 35 4 AIV 100 FPC-V-153 35 4 AIV (h) 100 FPC-V-154 35 4 AIV (h) 49 HPCS-RV-14 N/A N/A OCIV (g)(j) 49 HPCS-RV-35 N/A N/A OCIV (g)(j) 49 HPCS-V-12 N/A N/A OCIV (p)(k)(m) 49 HPCS-V-23 180 11 AIV (p) 6 HPCS-V-4 N/A N/A OCIV (i)(c)(k)(m) 58 HCV-V-1 N/A N/A OCIV 58 HCV-V-2 N/A N/A OCIV 6 HPCS-V-5 N/A N/A OCIV (i)(c) 31 HPCS-V-15 N/A N/A OCIV (c)(p)(k)(m) 78e HPCS-V-65 N/A N/A MCIV 78e HPCS-V-68 N/A N/A MCIV 63 LPCS-FCV-11 N/A N/A OCIV (p)(k)(m) 63 LPCS-V-12 180 10 AIV (p) 63 LPCS-RV-18 N/A N/A OCIV (g)(j) 63 LPCS-RV-31 N/A N/A OCIV (g)(j)

(a) See Technical Specification Bases 3.3.6.1 for the isolation signal(s) which operate each group.

(c) Valve leakage not included in sum of Type B and C tests.

(g) Not subject to Type C Leak Rate Test.

(h) Hydraulic leak test at 1.10 Pa.

(i) Not subject to Type C test. Test per Technical Specification SR 3.4.6.1.

(j) Tested as part of Type A test.

(k) Automatic open/close instrumentation supports ECCS operability; when not in use for ECCS normal position is closed.

(m) PCIV is operable (per TS 3.6.1.3) with non-functional open/close instrumentation if closed, secured and disabled from automatically opening.

(p) Not subject to leak rate testing (SR 3.6.1.1.1 and SR 3.6.1.3.12).

10 of 15

ILC-24 NRC Validation Exam Question: SRO-86 PCIVs 3.6.1.3 3.6 CONTAINMENT SYSTEMS 3.6.1.3 Primary Containment Isolation Valves (PCIVs)

LCO 3.6.1.3 Each PCIV, except reactor building-to-suppression chamber vacuum breakers, shall be OPERABLE.

APPLICABILITY: MODES 1, 2, and 3 ACTIONS

NOTES----------------------------------------------------------

1. Penetration flow paths may be unisolated intermittently under administrative controls.

2. Separate Condition entry is allowed for each penetration flow path.

3. Enter applicable Conditions and Required Actions for systems made inoperable by PCIVs.

4. Enter applicable Conditions and Required Actions of LCO 3.6.1.1, "Primary Containment,"

when PCIV leakage results in exceeding overall containment leakage rate acceptance criteria.

CONDITION REQUIRED ACTION COMPLETION TIME A. ------------NOTE------------ A.1 Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Only applicable to penetration flow path by main steam line penetration flow paths use of at least one closed with two PCIVs. and de-activated automatic AND

valve, closed manual valve, blind flange, or check valve 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for main One or more penetration with flow through the valve steam line flow paths with one secured.

PCIV inoperable for reasons other than AND Condition D.

Columbia Generating Station 3.6.1.3-1 Amendment No. 169,208 225 251 11 of 15

ILC-24 NRC Validation Exam Question: SRO-86 PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. (continued) A.2 ---------------NOTE--------------

1. Isolation devices in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.

Verify the affected Once per 31 days for penetration flow path is isolation devices isolated. outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment Columbia Generating Station 3.6.1.3-2 Amendment No. 169,208 225 12 of 15

ILC-24 NRC Validation Exam Question: SRO-86 PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME B. ------------NOTE------------ B.1 Isolate the affected 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Only applicable to penetration flow path by penetration flow paths use of at least one closed with two PCIVs. and de-activated automatic

valve, closed manual valve, or blind flange.

One or more penetration flow paths with two PCIVs inoperable for reasons other than Condition D.

C. ------------NOTE------------ C.1 Isolate the affected 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months except for Only applicable to penetration flow path by excess flow check penetration flow paths use of at least one closed valves (EFCVs) with only one PCIV. and de-activated automatic

valve, closed manual valve, AND or blind flange.

One or more penetration 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for EFCVs flow paths with one PCIV inoperable for AND reasons other than Condition D.

Columbia Generating Station 3.6.1.3-3 Amendment No. 169,208 225 13 of 15

ILC-24 NRC Validation Exam Question: SRO-86 PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. (continued) C.2 --------------NOTES-------------

1. Isolation devices in high radiation areas may be verified by use of administrative means.

2. Isolation devices that are locked, sealed, or otherwise secured may be verified by use of administrative means.

Verify the affected penetration flow path is Once per 31 days for isolated. isolation devices outside primary containment AND Prior to entering MODE 2 or 3 from MODE 4 if primary containment was de-inerted while in MODE 4, if not performed within the previous 92 days, for isolation devices inside primary containment Columbia Generating Station 3.6.1.3-4 Amendment No. 169,208 225 14 of 15

ILC-24 NRC Validation Exam Question: SRO-86 PCIVs 3.6.1.3 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME D. One or more secondary D.1 Restore leakage rate to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for containment bypass within limit. hydrostatically tested leakage rate, MSIV line leakage not on a leakage rate, or closed system hydrostatically tested lines leakage rate not AND within limit.

4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for secondary containment bypass leakage AND 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months for MSIV leakage AND 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months for hydrostatically tested line leakage on a closed system E. Required Action and E.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A, B, AND C, or D not met.

E.2 Be in MODE 4. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months Columbia Generating Station 3.6.1.3-5 Amendment No. 169,208 225 251 15 of 15

ILC-24 NRC Validation Exam Question: SRO-89 125 and 250 VDC Battery Parameters 1.8.6.2 1.8 ELECTRICAL POWER SYSTEMS 1.8.6.2 125 and 250 VDC Battery Parameters RFO 1.8.6.2 Battery parameters for the Division 1, 2, and 3, 125 and Division 1 250 VDC batteries shall be within the limits.

APPLICABILITY: When the associated DC electrical power systems are required to be OPERABLE.

COMPENSATORY MEASURES

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each battery.

REQUIRED CONDITION COMPENSATORY MEASURE COMPLETION TIME A. One or more batteries A.1 Verify float voltage 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months with float voltage not in 126/252 V.

range (129 to 132 / 258 to 264 V). AND A.2 Return float voltage to be 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months within range.

B. One or more batteries B.1 Verify electrolyte level 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months with electrolyte level not greater than or equal to the in range (> low level low level mark.

mark and 1/4" above high level mark). AND

NOTE---------------

Required Compensatory Measure B.2 is only applicable if electrolyte level was below the top of plates.

B.2 Equalize and perform a 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months service test.

AND Columbia Generating Station 1.8.6.2-1 Revision 73 1 of 4

ILC-24 NRC Validation Exam Question: SRO-89 125 and 250 VDC Battery Parameters 1.8.6.2 COMPENSATORY MEASURES REQUIRED CONDITION COMPENSATORY MEASURE COMPLETION TIME B. (continued). B.3 Return level to be within 31 days range.

C. One or more batteries C.1 Verify affected connection 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with corrosion identified. resistance is less than allowed.

AND C.2 Remove corrosion. 7 days D. One or more battery D.1 Verify room temperature is: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months rooms with ventilation not operating. a. 74°F for Division 1 and 2 batteries; and

b. 65°F for Division 3 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months battery.

AND D.2 Verify affected battery(s) are not on equalize.

E. One or more batteries E.1 Initiate Condition Report. Immediately with battery cell electrolyte temperature: AND

a. For Division 1 and E.2 Verify room temperature is: 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months 2 batteries, < 74°F; and a. 74°F for Division 1 and 2 batteries; and

b. For Division 3 battery < 65°F. b. 65°F for Division 3 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months battery.

AND E.3 Restore battery cell temperature within limit.

Columbia Generating Station 1.8.6.2-2 Revision 100 2 of 4

ILC-24 NRC Validation Exam Question: SRO-89 125 and 250 VDC Battery Parameters 1.8.6.2 COMPENSATORY MEASURES (continued)

REQUIRED CONDITION COMPENSATORY MEASURE COMPLETION TIME F. One or more batteries F.1 Verify remaining cell float 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with one or more cells voltages 2.07 V.

with individual cell float voltage < 2.13 V. AND F.2 Monitor subject cell voltage. Every 31 days G. One or more batteries G.1 Verify float current 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months with float current 2 amps.

> 1 amp.

AND G.2 Restore current to 1 amp. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months H. One or more batteries H.1 Verify float voltage 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months with individual cell 126/252 V.

specific gravity < 1.195 or battery average AND specific gravity 1.205.

H.2 Verify float current 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2 amps.

AND H.3 Verify all cell voltage 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months 2.07 V.

AND H.4 Restore specific gravity to 92 days be within limits.

Columbia Generating Station 1.8.6.2-3 Revision 73 3 of 4

ILC-24 NRC Validation Exam Question: SRO-89 125 and 250 VDC Battery Parameters 1.8.6.2 COMPENSATORY MEASURES (continued)

REQUIRED CONDITION COMPENSATORY MEASURE COMPLETION TIME I. One or more batteries I.1 Initiate a Condition Report. Immediately with connection resistance greater than allowed.

OR Cell crack or leakage, appearance or rack issues identified.

OR Required Compensatory Measure and associated Completion Time of Condition A, B, C, D, E, F, G, or H not met.

Columbia Generating Station 1.8.6.2-4 Revision 73 4 of 4

ILC-24 NRC Validation Exam Question: SRO-91 RPS Instrumentation 3.3.1.1 3.3 INSTRUMENTATION 3.3.1.1 Reactor Protection System (RPS) Instrumentation LCO 3.3.1.1 The RPS instrumentation for each Function in Table 3.3.1.1-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.1-1 ACTIONS

NOTE-----------------------------------------------------------

Separate Condition entry is allowed for each channel.

CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Place channel in trip. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months channels inoperable.

OR

NOTE---------------

Not applicable for Functions 2.a, 2.b, 2.c, 2.d, or 2.f.

A.2 Place associated trip 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months system in trip.

NOTE-------------- B.1 Place channel in one trip 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Not applicable for system in trip.

Functions 2.a, 2.b, 2.c, 2.d, or 2.f. OR B.2 Place one trip system in 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. One or more Functions trip.

with one or more required channels inoperable in both trip systems.

Columbia Generating Station 3.3.1.1-1 Amendment No. 169 225 226 253 1 of 4

ILC-24 NRC Validation Exam Question: SRO-91 RPS Instrumentation 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. One or more Functions C.1 Restore RPS trip capability. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months with RPS trip capability not maintained.

D. Required Action and D.1 Enter the Condition Immediately associated Completion referenced in Time of Condition A, B, Table 3.3.1.1-1 for the or C not met. channel.

E. As required by Required E.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Action D.1 and POWER to < 29.5% RTP.

referenced in Table 3.3.1.1-1.

F. As required by Required F.1 Be in MODE 2. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

G. As required by Required G.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and referenced in Table 3.3.1.1-1.

H. As required by Required H.1 Initiate action to fully insert Immediately Action D.1 and all insertable control rods in referenced in core cells containing one or Table 3.3.1.1-1. more fuel assemblies.

Columbia Generating Station 3.3.1.1-2 Amendment No. 169 225 226 253 2 of 4

ILC-24 NRC Validation Exam Question: SRO-91 RPS Instrumentation 3.3.1.1 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME I. As required by Required I.1 Initiate alternate method to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Action D.1 and detect and suppress referenced in thermal hydraulic instability Table 3.3.1.1-1. oscillations.

AND

NOTE-------------

LCO 3.0.4 is not applicable.

I.2 Restore required channels 120 days to OPERABLE.

J. Required Action and J.1 Reduce THERMAL 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months associated Completion POWER to less than the Time of Condition I not value specified in the met. COLR.

SURVEILLANCE REQUIREMENTS

NOTES----------------------------------------------------------

1. Refer to Table 3.3.1.1-1 to determine which SRs apply for each RPS Function.

2. When a channel is placed in an inoperable status solely for performance of required Surveillances, entry into associated Conditions and Required Actions may be delayed for up to 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months provided the associated Function maintains RPS trip capability.

SURVEILLANCE FREQUENCY SR 3.3.1.1.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.1.1-3 Amendment No. 169 225 226 253 3 of 4

ILC-24 NRC Validation Exam Question: SRO-91 RPS Instrumentation 3.3.1.1 Table 3.3.1.1-1 (page 3 of 4)

Reactor Protection System Instrumentation APPLICABLE CONDITIONS MODES OR REQUIRED REFERENCED OTHER CHANNELS FROM SPECIFIED PER TRIP REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS SYSTEM ACTION D.1 REQUIREMENTS VALUE

3. Reactor Vessel Steam 1,2 2 G SR 3.3.1.1.8 1079 psig Dome Pressure - High SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

4. Reactor Vessel Water 1,2 2 G SR 3.3.1.1.1 9.5 inches Level - Low, Level 3 SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

5. Main Steam Isolation Valve 1 8 F SR 3.3.1.1.8 12.5% closed

- Closure SR 3.3.1.1.10 SR 3.3.1.1.14 SR 3.3.1.1.15

6. Primary Containment 1,2 2 G SR 3.3.1.1.8 1.88 psig Pressure - High SR 3.3.1.1.10 SR 3.3.1.1.14

7. Scram Discharge Volume Water Level - High

a. Transmitter/Level 1,2 2 G SR 3.3.1.1.1 529 ft 9 inches Indicating Switch SR 3.3.1.1.8 elevation SR 3.3.1.1.10 SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.1 529 ft 9 inches SR 3.3.1.1.8 elevation SR 3.3.1.1.10 SR 3.3.1.1.14

b. Transmitter/Level 1,2 2 G SR 3.3.1.1.8 529 ft 9 inches Switch SR 3.3.1.1.10(d)(e) elevation SR 3.3.1.1.14 5(a) 2 H SR 3.3.1.1.8 529 ft 9 inches SR 3.3.1.1.10(d)(e) elevation SR 3.3.1.1.14 (a) With any control rod withdrawn from a core cell containing one or more fuel assemblies.

(d) If the as-found channel setpoint is outside its predefinded as-found tolerance, then the channel shall be evaluated to verify that it is functioning as required before returning the channel to service.

(e) The instrument channel setpoint shall be reset to a value that is within the as-left tolerance around the Limiting Trip Setpoint (LTSP) at the completion of the surveillance; otherwise, the channel shall be declared inoperable.

Setpoints more conservative than the LTSP are acceptable provided that the as-found and as-left tolerances apply to the actual setpoint implemented in the surveillance procedures (Nominal Trip Setpoint) to confirm channel performance. The LTSP and the methodologies used to determine the as-found and the as-left tolerances are specified in the Licensee Controlled Specifications.

Columbia Generating Station 3.3.1.1-9 Amendment No.169 225 226 253 4 of 4

ILC-24 NRC Validation Exam Question: SRO-97 SRM Instrumentation 3.3.1.2 3.3 INSTRUMENTATION 3.3.1.2 Source Range Monitor (SRM) Instrumentation LCO 3.3.1.2 The SRM instrumentation in Table 3.3.1.2-1 shall be OPERABLE.

APPLICABILITY: According to Table 3.3.1.2-1.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. One or more required A.1 Restore required SRMs to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months SRMs inoperable in OPERABLE status.

MODE 2 with intermediate range monitors (IRMs) on Range 2 or below.

B. Three required SRMs B.1 Suspend control rod Immediately inoperable in MODE 2 withdrawal.

with IRMs on Range 2 or below.

C. Required Action and C.1 Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months associated Completion Time of Condition A or B not met.

D. One or more required D.1 Fully insert all insertable 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SRMs inoperable in control rods.

MODE 3 or 4.

AND D.2 Place reactor mode switch 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months in the shutdown position.

Columbia Generating Station 3.3.1.2-1 Amendment No. 149,169 225 1 of 3

ILC-24 NRC Validation Exam Question: SRO-97 SRM Instrumentation 3.3.1.2 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME E. One or more required E.1 Suspend CORE Immediately SRMs inoperable in ALTERATIONS except for MODE 5. control rod insertion.

AND E.2 Initiate action to fully insert Immediately all insertable control rods in core cells containing one or more fuel assemblies.

SURVEILLANCE REQUIREMENTS

NOTE-----------------------------------------------------------

Refer to Table 3.3.1.2-1 to determine which SRs apply for each applicable MODE or other specified conditions.

SURVEILLANCE FREQUENCY SR 3.3.1.2.1 Perform CHANNEL CHECK. In accordance with the Surveillance Frequency Control Program Columbia Generating Station 3.3.1.2-2 Amendment No. 149,169 225 238 2 of 3

ILC-24 NRC Validation Exam Question: SRO-97 SRM Instrumentation 3.3.1.2 Table 3.3.1.2-1 (page 1 of 1)

Source Range Monitor Instrumentation APPLICABLE MODES OR OTHER SPECIFIED REQUIRED SURVEILLANCE FUNCTION CONDITIONS CHANNELS REQUIREMENTS (a)

1. Source Range Monitor 2 3 SR 3.3.1.2.1 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7 3, 4 2 SR 3.3.1.2.3 SR 3.3.1.2.4 SR 3.3.1.2.6 SR 3.3.1.2.7 5 2 (b), (c) SR 3.3.1.2.1 SR 3.3.1.2.2 SR 3.3.1.2.4 SR 3.3.1.2.5 SR 3.3.1.2.7 (a) With IRMs on Range 2 or below.

(b) Only one SRM channel is required to be OPERABLE during spiral offload or reload when the fueled region includes only that SRM detector.

(c) Special movable detectors may be used in place of SRMs if connected to normal SRM circuits.

Columbia Generating Station 3.3.1.2-5 Amendment No. 149,169 225 3 of 3

2021 NRC Exam Question: SRO-100 Page 1 of 1

ML21113A005

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