ML18129A044

2018 Fermi Ile As-Administered Written Exam with Provided References

ML18129A044
Person / Time
Site:	Fermi
Issue date:	05/07/2018
From:	Chuck Zoia NRC/RGN-III
To:	Detroit Edison, Co
Zoia C
Shared Package
ML17164A400	List: ML18128A312 ML18128A314 ML18129A033 ML18129A044
References
Download: ML18129A044 (370)

Text

NOTE: Per NUREG 1021 ES-501 C.1.B this exam key has been annotated to indicate changes made while administering the exam.

Added text is in RED Removed text is lined out Questions changed:

Question: R31 Comment: Are A&C CMCs in RUN because of Surveillance or in Auto for Surveillance?

Action(s) taken: After conferring with the chief examiner, the following was added to the stem of the question: running for Division 1 Core Spray System Pump and Valve Operability Test.

Question: S88 Comment: Candidate started to ask a question about Q S88. Candidate asked What MODE is the plant in?

Action(s) taken: After conferring with the chief examiner, the candidates were informed that the plant is in MODE 1.

Question: R42 Comment: Is the question stating that it is a loss of ALL Offsite Power?

Action(s) taken: After conferring with the chief examiner, informed candidates that the question is referring to a loss of ALL Offsite Power.

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 1 of 282 Question 1

1 K/A Importance: 3.5/3.6 Points: 1.00 R01 Difficulty: 3.00 Level of Knowledge: High Source: BANK 73969 The plant was at rated conditions when a Main Turbine Trip and ATWS occurred.

Given the following:

• SLC is being injected.

• Reactor power is 22% and slowly lowering.

• RPV water level is 50 and intentionally being lowered.

• Lo Lo-SET SRVs are lifting and Torus temperature is 112°F and rising.

For the conditions above; which of the following is a valid reason for continuing to lower RPV water level?

A. To prevent thermal-hydraulic instabilities.

B. To further reduce feedwater inlet subcooling.

C. To reduce natural circulation and reactor power.

D. To lower the amount of voids (void fraction) within the core.

Answer: C ILO 2018 NRC Exam Page: 2 of 282 Question 1

Answer Explanation:

As RPV water level decreases, the height of the fluid columns is reduced, thereby reducing the natural circulation driving head. This lowers the natural circulation flow which raises the void fraction, thereby adding negative reactivity and lowering reactor power.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Subcooling is minimized below 114, therefore, thermal-hydraulic instabilities will be prevented.

B. With RPV level already below 114, the feedwater spargers will be uncovered and further reduction in subcooling will not occur.

D. Lowering RPV level will lower natural circulation which will raise the amount of voids in the core causing void fraction to rise.

Reference Information:

Appendix B-Technical Basis, Volume 1 ST-OP-802-3003-001 Student Texts EOP RPV Control NUREG 1123 KA Catalog Rev. 2 295001 AK1 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION :

295001 AK1.01 Natural circulation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK High RO Associated objective(s):

RPV Control Cycle 12-3 Objectives Exam Objectives Cognitive Terminal Describe the expected plant/system response for actions directed by 29.100.01 Sh 1, 1A, 3, and 3A.

ILO 2018 NRC Exam Page: 3 of 282 Question 1

2 K/A Importance: 4.1/4.2 Points: 1.00 R02 Difficulty: 3.00 Level of Knowledge: Low Source: BANK 73987 If an EDG were already running in parallel with Offsite Power, how will the EDG respond to a loss of Offsite Power?

A. The EDG output breaker will trip and the EDG will shut down.

B. The EDG output breaker will trip, Load Shed will occur, and the EDG output breaker will reclose.

C. The EDG output breaker will remain closed, the EDG will shut down, and then it will restart in isochronous mode.

D. The EDG output breaker will remain closed, the EDG will continue running and the its governor will shift to isochronous mode.

Answer: B ILO 2018 NRC Exam Page: 4 of 282 Question 2

Answer Explanation:

When the EDG is operating in parallel with offsite power, the undervoltage relays are bypassed by a contact in the closed EDG output breaker. If an actual loss of power to the bus from offsite occurs the EDG under frequency relay will open (TRIP) the EDG output breaker and the undervoltage scheme will operate as designed (load shed, close output breaker).

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Although the output breaker will trip, the EDG will not shut down.

C. The EDG will remain running, and output breaker will trip. The EDG will be in isochronous, but does not restart.

D. Output breaker trips, other parts are correct.

Reference Information:

ST-OP-315-0065 EDG Student Text i-n-2711-36 EDG 14 Control Drawing NUREG 1123 KA Catalog Rev. 2 295003 AK2 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C.

POWER and the following:

295003 AK2.02 Emergency generators 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK Low RO Associated objective(s):

Emergency Diesel Generator (R3000)

Cognitive Terminal Given various controls and indications for Emergency Diesel Generator operations, analyze that indication to determine proper plant response in accordance with approved plant procedures.

ILO 2018 NRC Exam Page: 5 of 282 Question 2

3 K/A Importance: 3.1/3.5 Points: 1.00 R03 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 74089 A loss of which DC electrical distribution cabinet would result in a Reactor Scram at full power and why?

A. 2PC3-Main Turbine Trip B. 2IA-INOP trip of SRMs/IRMs C. 2PB2-Outboard MSIV closure D. 2PA2-Loss of Scram Pilot Air Header Answer: A ILO 2018 NRC Exam Page: 6 of 282 Question 3

Answer Explanation:

Loss of BOP DC 2PC3 will cause a Main Turbine trip and subsequent reactor scram if power is greater than 30%.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. Loss of 48/24VDC power would generate an INOP trip for the SRMs and IRMs but would not cause a Reactor Scram at full power.

C. Loss of DC power would cause a loss of MSIV position indication and possible MSIV closure if MSIV AC solenoids are deenergized. However, under the conditions stated in the stem, the MSIV AC solenoids will be energized, so de-energization of this DC source will not result in MSIV closure under the given plant conditions.

D. Loss of Division 1 and/or 2 ESF DC would prevent operation of the RPS Backup Scram Valves.

Reference Information:

23.309 260/130 V DC Electrical System ST0064001 NUREG 1123 KA Catalog Rev. 2 295004 AK3. Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER :

295004 AK3.03 3.1/3.5 Reactor SCRAM: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

DC Electrical Distribution (R3200 & S3102)

Cognitive Enabler Discuss the DC Electrical Distribution System interrelationships with other systems.

ILO 2018 NRC Exam Page: 7 of 282 Question 3

4 K/A Importance: 3.6/3.6 Points: 1.00 R04 Difficulty: 2.00 Level of Knowledge: Low Source: BANK 73991 Following a trip of the Main Generator from 25% power, which of the following represents the expected status of the reactor/turbine pressure regulating system?

Turbine Stop Valves Turbine Control Valves Turbine Bypass Valves A. Open Open Closed B. Closed Open Open C. Closed Closed Open D. Open Closed Closed Answer: C ILO 2018 NRC Exam Page: 8 of 282 Question 4

Answer Explanation:

A Main Generator trip will result in a trip of the Main Turbine. In order to protect the Main Turbine, the Turbine control valves and Turbine stop valves will close to prevent steam from entering the main turbine.

The Turbine Bypass Valves will open and modulate to control Reactor pressure at the pressure regulator setpoint.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could fail to understand that a Generator Trip causes a Main Turbine trip at 25% and incorrectly think the bypass valves would stay closed since the Turbine stop valves are listed as open.

Closure of the stop or control valves would prevent steam addmitance to the turbine and result in bypass valves opening to control pressure C. The candidate could fail to understand that a Generator Trip causes a Main Turbine trip and incorrectly assume the control and stop valves would stay open. This would result in the reactor/turbine pressure systems not needing the bypass valves to open to control reactor pressure D. The candidate could incorrectly assume that the turbine control valves stay open on a turbine trip with bypass valves opening to control reactor/turbine pressure.

Reference Information:

23.109 Turbine Operating Procedure NUREG 1123 KA Catalog Rev. 2 295005 AA1. Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP:

295005 AA1.05 3.6/3.6 Reactor/turbine pressure regulating system 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK Low RO Associated objective(s):

Main Generator and Excitation (N3031 & N3034)

Cognitive Enabler Discuss the Main Generator and Excitation System interrelationships with other systems.

ILO 2018 NRC Exam Page: 9 of 282 Question 4

5 K/A Importance: 3.5/3.8 Points: 1.00 R05 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 73995 Which of the following inputs would directly cause a Reactor Scram?

A. Condenser Vacuum 3.68 psia.

B. Wide Range Level indication of 173".

C. 3 Main Steam Isolation Valves close.

D. Closure of Turbine Stop Valves 1 and 3.

Answer: C ILO 2018 NRC Exam Page: 10 of 282 Question 5

Answer Explanation:

MSIV closure logic is such that closure of any 1 MSIV will not cause a direct RPS actuation on either side.

Closure of 2 MSIVs (A and B, C and D for RPS A; A and C or B and D for RPS B) may cause a 1/2 scram (Note: Closure of MSIVs B and C or A and D will NOT cause a half scram on either RPS trip system).

Closure of any 3, of the 4, MSIVs will always cause a full reactor scram.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Vacuum of 3.68 will cause a turbine trip which will indirectly cause a Scram B. The Scram signal for level comes from Narrow Range Instruments.

D. Closure of TSV 1&3 will cause a half Scram, not a full scram.

Reference Information:

ARP 3D73 3D74 NUREG 1123 KA Catalog Rev. 2 295006 AA2. Ability to determine and/or interpret the following as they apply to SCRAM :

295006 AA2.06 3.5/3.8 Cause of reactor SCRAM 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Reactor Protection System (C7100)

Cognitive Enabler List the automatic features of Reactor Protection System operations.

ILO 2018 NRC Exam Page: 11 of 282 Question 5

6 K/A Importance: 4.3/4.4 Points: 1.00 R06 Difficulty: 4.00 Level of Knowledge: High Source: NEW 74626 The plant is operating at 68% power when the Control Room is abandoned due to toxic fumes. A reactor scram could NOT be completed prior to abandonment. The ENTIRE Control Room Envelope is inaccessible due to the toxic fumes.

Which of the following is the method of performing a reactor scram for the given plant conditions?

A. Trip Turbine Trip Relay TTR1 or TTR2.

B. Take two operable APRM Mode switches out of Operate.

C. Open circuit breakers CB2A (C71-P001A) and CB2B (C71-P001B).

D. Place Main Steam Line Radiation Monitors (A or C) and (B or D) in INOP.

Answer: C ILO 2018 NRC Exam Page: 12 of 282 Question 6

Answer Explanation:

Control Room Abandonment requires the crew to enter AOP 20.000.19, Shutdown from Outside the Main Control Room. AOP 20.000.19 requires the crew to attempt to scram the reactor from the Main Control Room and then proceed to H21-P100, Remote Shutdown Panel. The Remote Shutdown Panel does not include reactor shutdown and turbine trip controls because it is assumed that reactor shutdown and turbine trip will be performed prior to leaving the Control Room.

If this action is not possible, then reactor shutdown will be initiated by taking two APRMs out of operation per Condition B of 20.000.19, Shutdown from Outside the Control Room. However, this action is also not possible because the actions are taken in the Relay Room which is inside the Control Room Envelope, which is inaccessible as stated in the stem of the question.

Therefore, the candidate should determine that the actions of 20.000.19, Condition C are required, which will have the operator scram the reactor by opening CB2A and CB2B.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Tripping Turbine Trip Relay TTR1 or TTR2 is an action from 20.000.19 and would result in a reactor scram above 30% power. However, these actions are only necessary to trip the Main Turbine and should be performed after the reactor is scrammed. Additionally, these actions are performed in the Relay Room thereby requiring entry into the Control Room Envelope, which is NOT accessible.

B. Taking 2 operable APRMs out of operate is an action from 20.000.19 and would result in a reactor scram. However, these actions are performed in the Relay Room thereby requiring entry into the Control Room Envelope, which is NOT accessible.

D. Placing two Main Steam Line Radiation Monitors out of Operate (in the logic sequence specified in the distractor) is an action taken from 20.000.19 and would result in a direct reactor scram. However, these actions are performed, when Reactor Pressure is lowering too fast, in an attempt to lower the cooldown rate by closing the MSIVs (by initiating an NSSSS Group 1 Isolation). Additionally, these actions are performed in the Relay Room thereby requiring entry into the Control Room Envelope, which is NOT accessible.

Reference Information:

20.000.19, Shutdown from Outside the Main Control Room 20.000.19 BASES, Shutdown from Outside the Main Control Room Bases.

NUREG 1123 KA Catalog Rev. 2 295016 Control Room Abandonment G2.1.23 Ability to perform specific system and integrated plant procedures during different modes of plant operation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Reactor Operator Performance Enabler Perform proper system operations in accordance with System Operating Procedures (SOP).

ILO 2018 NRC Exam Page: 13 of 282 Question 6

7 K/A Importance: 3.2/3.5 Points: 1.00 R07 Difficulty: 3.00 Level of Knowledge: High Source: BANK 74004 The plant is operating at 65% with P4300-C001 (North) and P4300-C002 (Center) TBCCW Pumps in service, when the following occurs:

• 5D5, TBCCW PUMPS DIFFERENTIAL PRESSURE HIGH/LOW, alarms

• P43-R805, TBCCW Headers Pressure Indicator, indicates a d/p of 31 psig.

With these conditions, which one of the following is the correct diagnosis and required action?

A. TBCCW Differential Pressure is too HIGH; it is required to STOP an operating TBCCW Pump.

B. TBCCW Differential Pressure is too LOW; it is required to REDUCE Main Turbine Generator Load.

C. TBCCW Differential Pressure is too LOW; it is required to START P4300-C003 (South) TBCCW Pump.

D. TBCCW Differential Pressure is too HIGH; it is required to CLOSE P43-F405, TBCCW DP Control Valve.

Answer: A ILO 2018 NRC Exam Page: 14 of 282 Question 7

Answer Explanation:

The purpose of the TBCCW Pressure Control Valve (PCV) is to maintain a constant differential pressure between the supply and return headers of the system, thereby supplying the required amount of flow to all loads. The PCV operates in response to pressure signals to maintain a differential pressure of 20 to 30 psid between the supply and return headers. The PCV will respond to system demand by heat load TCVs. With the Pressure Indicator at 31 psig, TBCCW DP is too high. With pressure too high it is required to shut down an operating pump as directed by ARP 5D5 Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could incorrectly assume that a DP of 31 psig is too low and perform actions to reduce load on the MTG. This action would result in the PCV opening to adjust for the flow and keep pressure the same, this action is only performed if TBCCW temperature cannot be maintained.

C. The candidate could incorrectly assume that a DP of 31 psig is too low and perform actions to start the standby TBCCW pump. Starting an additional pump would cause the PCV to open and result in raising the DP.

D. The candidate could incorrectly assume that closing the P43-F405 TBCCW PCV would lower the DP.

Closing the P43-F405 TBCCW DP PCV would result in an increase in TBCCW DP.

Reference Information:

ARP 5D5 NUREG 1123 KA Catalog Rev. 2 295018 AA2 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER :

295018 AA2.03 3.2/3.5 Cause for partial or complete loss 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

BANK High RO Associated objective(s):

Turbine Building Closed Cooling Water (P4300)

Cognitive Terminal Given various controls and indications for Turbine Building Closed Cooling Water operations, analyze that indication to determine proper plant response in accordance with approved plant procedures.

ILO 2018 NRC Exam Page: 15 of 282 Question 7

8 K/A Importance: 3.5/3.4 Points: 1.00 R08 Difficulty: 3.00 Level of Knowledge: Low Source: BANK 74686 The plant is operating at 100% power.

• Annunciator 7D53, Station Air Header Pressure Low, Alarms.

• P50-R802, Station Air Header Pressure is 90 psig, lowering.

The FIRST required action per 20.129.01, Loss of Station and / or Control Air is ___(1)___.

The BASIS for this action is to avoid conditions which will result in ___(2)___ MSIVs drifting SHUT.

A. (1) Start any available Station Air Compressors (2) Inboard B. (1) Start any available Station Air Compressors (2) Outboard C. (1) Start Div I and Div II Control Air Compressors (2) Inboard D. (1) Start Div I and Div II Control Air Compressors (2) Outboard Answer: B ILO 2018 NRC Exam Page: 16 of 282 Question 8

Answer Explanation:

20.129.01, Loss of Station and Control Air Condition A contains the guidance for the actions in the correct answer. 20.129.01 BASES contains the reasons for taking these actions. Specifically the basis for Caution 1, which is applicable for Condition A, is what is being tested in this question to determine if the operator knows the reason for the operation of the standby Station Air Compressor.

The Candidate must know that the Station Air Compressors are started prior to Div I/Div II Control Air Compressors. The Station Air Compressor aligned for standby should have AUTO started at 95 psig. The THIRD Station Air Compressor has no auto start and requires manual action to start, per the AOP, when Station Air Pressure reaches 90 psig. The third Station Air Compressor is considered an available Station Air Compressor.

The Candidate must also know that starting the standby Station Air Compressor will stabilize the Station Air and Interruptible Air headers, thereby preventing the impacts listed in CAUTION 1. The candidate must recall that the OUTBOARD MSIVs drift closed on a loss of Interruptible Air pressure. Note: The INBOARD MSIVs have Nitrogen pneumatic supply with air backup from Division 1 NIAS.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could correctly recall that the FIRST required action of the AOP is to start any available Station Air Compressors. However, the candidate could incorrectly recall that the reason for doing so is to prevent closure of the Inboard MSIVs, which is incorrect because the Inboard MSIVs are supplied by Nitrogen and Nitrogen accumulators with Division 1 NIAS as a backup supply.

C. The candidate could incorrectly recall that the FIRST required action of the AOP is to start the Div 1 and Div 2 Control Air Compressors, which is incorrect since the Control Air Compressors do not auto start until Station Air pressure drops to 85 psig, per Condition B of the AOP. The Station Air Compressors should be started first in an attempt to stabilize the air system, preventing further degradation and automatic system actuations. This could lead the candidate to also conclude that the Control Air Compressors are started in order to prevent closure of the Inboard MSIVs, which is plausible because of the relationship between the MSIVs and Div 1 NIAS, which is a backup to the normal Nitrogen pneumatic supply inside the Drywell.

D. The candidate could incorrectly recall that the FIRST required action of the AOP is to start the Div 1 and Div 2 Control Air Compressors, which is incorrect since the Control Air Compressors do not auto start until Station Air pressure drops to 85 psig, per Condition B of the AOP. The Station Air Compressors should be started first in an attempt to stabilize the air system, preventing further degradation and automatic system actuations.

Reference Information:

20.129.01, Loss of Station and/or Control Air 20.129.01 BASES, Loss of Station and/or Control Air Bases NUREG 1123 KA Catalog Rev. 2 295019 Partial or Complete Loss of Instrument Air 295019 AK3. Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:

295019 AK3.02 3.5/3.4 Standby air compressor operation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

BANK Low RO ILO 2018 NRC Exam Page: 17 of 282 Question 8

Associated objective(s):

Reactor Operator Performance Enabler Apply notes and cautions as directed by the EOP / AOP.

Reactor Operator Performance Enabler Perform other EOP / AOP actions per site procedures as directed.

ILO 2018 NRC Exam Page: 18 of 282 Question 8

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 19 of 282 Question 9

9 K/A Importance: 3.5/3.5 Points: 1.00 R09 Difficulty: 4.00 Level of Knowledge: High Source: MODIFIED 74746 The plant is in MODE 4 with Division 1 Residual Heat Removal (RHR) System in Shutdown Cooling (SDC) and RHR Pump A running. A valving error results in RPV water level lowering to 150" before it is corrected and the RPV level reduction is stopped.

In addition to depressing the E1150-F015A(B), Div 1(2) LPCI Inbd Iso Vlv, SDC Isolation RESET pushbuttons, of the numbered actions listed below, which are required to restore Shutdown Cooling?

1. Restore RPV level to 173 - 214".

2. Restore RPV level to 220 - 255".

3. Depress Inboard and Outboard MSIV Isolation Logic RESET pushbuttons.

4. Depress Hi Drywell Press & Reactor Low Level Logic RESET pushbuttons.

5. Depress Leak Detection Line Break Logic RESET pushbuttons.

A. 2 and 3 B. 1 and 4 C. 2, 3 and 4 D. 1, 4 and 5 Answer: A ILO 2018 NRC Exam Page: 20 of 282 Question 9

Answer Explanation:

The actions listed are from the Loss of Shutdown Cooling AOP, 20.205.01, Condition C for a Loss of SDC Condition caused by RPV Level dropping less than Level 3 (173"). The candidate must recognize the need to perform the following actions, for the reasons listed:

2. Restore RPV Level to the SDC level band (>220") to promote natural circulation within the RPV by raising RPV level above the bottom of the Steam Separators.

3. Depress Inboard and Outboard MSIV Isolation Logic RESET pushbuttons to reset the Group 4 (RHR SDC) isolation that occurred when RPV level dropped <L3, which will allow the SDC isolation valves to be re-opened.

These actions stabilize water level, reset the isolation signals and return RHR to service. Low reactor water Level 3 is selected as the isolation level because all normal operations of the RHR SDC mode should be performed with reactor water level above Level 3. If reactor water level decreases to Level 3, the RHR System is automatically isolated to isolate potential leakage paths in the RHR System that may result in draining of the reactor vessel water inventory. If the system isolates on decreasing level before LPCI initiates (at RPV Level 2, or 110.8"), the common SDC suction (E11-F008 and F009) valves and the LPCI injection valves (E11-F015A and F015B) in both divisions close, and pumps in the loop that is operating in the SDC mode trip on loss of suction path. The LPCI loop pumps that are lined up in standby mode are not affected.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could incorrectly determine that RPV Water level should be restored to the 173-214" level band per Action 1, which is plausible because this is the normal (non-Shutdown Cooling) shutdown RPV level band. The candidate could also determine that RHR initiation logic must be reset, per Action 4, by depressing the Hi Drywell Press & Reactor Low Level Logic RESET pushbuttons, which is plausible because a low RPV Water Level (<Level 3) condition has occurred, but incorrect because LPCI initiation occurs at Level 1.

C. The candidate could correctly determine that actions 2 and 3 need to be performed, however incorrectly conclude that Action 4 also needs to be performed to reset RHR initiation logic by depressing the Hi Drywell Press & Reactor Low Level Logic RESET pushbuttons, which is plausible because a low RPV Water Level (<Level 3) condition has occurred, but incorrect because LPCI initiation occurs at Level 1.

D. The candidate could incorrectly determine that RPV Water level should be restored to the 173-214" level band per Action 1, which is plausible because this is the normal (non-Shutdown Cooling) shutdown RPV level band. The candidate could also determine that RHR initiation logic must be reset, per Action 4, by depressing the Hi Drywell Press & Reactor Low Level Logic RESET pushbuttons, which is plausible because a low RPV Water Level (<Level 3) condition has occurred, but incorrect because LPCI initiation occurs at Level 1. The candidate could also determine that RHR leak detection line break logic needs reset, which is plausible because a low RPV Water Level (<Level 3) condition has occurred, but incorrect because RHR Leak Detection Line Break (LPCI Loop Select)

Logic actuates at RPV Level 2.

Reference Information:

20.205.01, Loss of Shutdown Cooling AOP.

NUREG 1123 KA Catalog Rev. 2 295021 Loss of Shutdown Cooling 295021 AA1.02 3.5/3.5 RHR/shutdown cooling 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High Low MODIFIED RO ILO 2018 NRC Exam Page: 21 of 282 Question 9

Associated objective(s):

Residual Heat Removal (E1100)

Cognitive Enabler List the interlocks associated with RHR System components.

Residual Heat Removal (E1100)

Cognitive Enabler List the interlocks associated with RHR System components.

ILO 2018 NRC Exam Page: 22 of 282 Question 9

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 23 of 282 Question 10

10 K/A Importance: 3.7/3.8 Points: 1.00 R10 Difficulty: 3.00 Level of Knowledge: High Source: BANK 74270 An irradiated fuel bundle is being removed from the core. An adjacent bundle has been lifted along with the selected bundle. When noticed, fuel movement is stopped, and the adjacent bundle falls back into the core.

When some bubbles come to the pool surface, the local Continuous Air Monitor (CAM) alarms.

D21-K717, RB5 Refuel Floor Lo Range ARM Ind trip Unit, indicates 10 mR/hr and slowly rising.

(1) Due to the trip of D21-K717 what automatic actions, if any, will occur AND (2) as a result of this event which of the following procedure(s) entry conditions is (are) met?

A. (1) RBHVAC isolates.

(2) 20.000.02, Abnormal Release of Radioactive Material, ONLY B. (1) RBHVAC isolates.

(2) 20.710.01, Refueling Floor High Radiation, AND 20.000.02, Abnormal Release of Radioactive Material C. (1) No automatic actions occur.

(2) 20.710.01, Refueling Floor High Radiation, AND 20.000.02, Abnormal Release of Radioactive Material D. (1) No automatic actions occur.

(2) 20.710.01, Refueling Floor High Radiation, AND 29.100.01, Sheet 5, Secondary Containment Control Answer: C ILO 2018 NRC Exam Page: 24 of 282 Question 10

Answer Explanation:

Per 16D1, if Rad Monitor Channel 15, 17 or 18 are verified greater than alarm setpoint, immediately evacuate the area, enter 20.710.01, Refueling Floor High Radiation, and enter 20.000.02, Abnormal Release Of Radioactive Material.

Distractor Explanation:

Distractors are plausible and incorrect because:

RBHVAC does not isolate based on the stem 29.100.01, Sheet 5, Secondary Containment Control, EOP entry conditions have not been met.

Reference Information:

20.710.01 (Entry Conditions) 16D1 (Actions)

Plant Procedures 16D01 NUREG 1123 KA Catalog Rev. 2 288000 A2. Ability to (a) predict the impacts of the following on the PLANT VENTILATION SYSTEMS

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations

288000 A2.04 High radiation: Plant-Specific 295023 AK2. Knowledge of the interrelations between REFUELING ACCIDENTS and the following:

295023 AK2.05 Secondary containment ventilation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Exam Usage LOR 2013 Exam NRC Question Use (ILO 2018)

BANK High RO Associated objective(s):

Cycle 12-1 Objectives Performance Objectives Performance Terminal Upon recognizing the entry conditions for a high radiation on the refueling floor, operators respond to control critical parameters within expected limits IAW approved plant procedures within validated completion times ILO 2018 NRC Exam Page: 25 of 282 Question 10

11 K/A Importance: 2.9/3.1 Points: 1.00 R11 Difficulty: 4.00 Level of Knowledge: Low Source: NEW 74606 The Torus Water Management System (TWMS) is in the TORUS FILL mode of operation.

The Torus is currently filling at 500 gpm via the Core Spray Test Line.

A LOCA subsequently occurs that causes a High Drywell Pressure condition.

Which of the following completes the statement below describing the impact of the High Drywell Pressure condition on Suppression Pool makeup?

Suppression Pool makeup will stop because _______

A. the North (South) Torus Water Management Pumps will trip.

B. G5100-F611, TWMS Cond to Torus Makeup Valve will close.

C. G5100-F612 (F613), TWMS Sec Cntm Inbd (Otbd) Iso Valves will close.

D. G5100-F606 (F607), TWMS Rtrn to CS Inbd (Otbd) Iso Valves will close.

Answer: D ILO 2018 NRC Exam Page: 26 of 282 Question 11

Answer Explanation:

The G5100-F606 (607), TWMS to CS Test Line Isolation Valves allow isolation of TWMS from the CS Test Line. When in the Torus Fill mode, these valves would be open. These valves are part of the Primary Containment Isolation System (PCIS) Group 12. They close upon actuation of Group 12 logic due to either High Drywell Pressure, Low RPV Water Level (Level 2), Drywell Floor Drain sump level high-high or Reactor Building Torus sump level high-high.

Therefore, these valves will go closed upon the high Drywell pressure condition specified in the stem of the question, which is what will cause Suppression Pool makeup to be lost.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could recall that the TWMS pumps would trip, if in another TWMS mode such as Torus Fill or Torus Cleanup, due to closure of the TWMS Pump Suction Valves on a Group 12 isolation signal. This answer is incorrect, however, because the TWMS pumps do NOT run when in the system is operating in the Torus Fill mode because the fill source comes from the Condensate System.

B. The candidate could recall that the G51-F611 is in the flow path that allows clean condensate, from the Condensate Filter Demineralizers, to be sent to the Torus. The candidate could incorrectly determine that this valve would go closed, to stop this flowpath, on a High Drywell Pressure condition.

This answer is incorrect, however, since the F611 only goes closed on a High/High Torus water level condition to prevent over-filling the Torus.

C. The candidate could recall that the G51-F612 (F613) are in the flow path that allows clean condensate, from the Condensate Filter Demineralizers, to be sent to the Torus. The candidate could assume that, since they are labeled as Secondary Containment Isolation Valves, they would go closed on a High Drywell Pressure condition, which is a common misconception among licensed operator candidates. However, although the F612/F613 form part of the secondary containment, they do NOT receive an automatic closure signal on High Drywell Pressure as most other Secondary Containment Isolation Valves do.

Reference Information:

23.144, TWMS System SOP M-5713, TWMS System FOS ST-OP-315-0069, TWMS System Student Text.

NUREG 1123 KA Catalog Rev. 2 295024 High Drywell Pressure.

295024 EK2.09 2.9/3.1 Suppression pool makeup: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Torus Water Management System (G5100)

Cognitive Enabler List the automatic features of Torus Water Management System operations.

Torus Water Management System (G5100)

Cognitive Enabler List the automatic features of Torus Water Management System operations.

ILO 2018 NRC Exam Page: 27 of 282 Question 11

12 K/A Importance: 4.4*/4.7* Points: 1.00 R12 Difficulty: 4.00 Level of Knowledge: High Source: BANK 74273 Consider each of the conditions listed below:

A - Reactor Pressure is 90 psi above the normal pressure regulator setpoint.

B - Reactor Pressure is at the Reactor Pressure Scram setpoint.

C - Reactor Pressure is 100 psi above the highest SRV safety setpoint.

D - Reactor Pressure is 175 psi above the highest SRV safety setpoint.

E - Reactor Pressure is 225 psi above the highest SRV safety setpoint.

Based on the HIGHEST Reactor Pressure encountered during each condition, which of these would have resulted in the plant NOT being in compliance with (1) LCO 3.4.11, Reactor Steam Dome Pressure, and/or (2) SL 2.1.2, Reactor Coolant System Pressure Safety Limit?

A. (1) ALL of the listed conditions (2) Conditions D and E only B. (1) Conditions B through E only (2) Conditions D and E only C. (1) ALL of the listed conditions (2) Conditions C through E only D. (1) Conditions B through E only (2) Conditions C through E only Answer: B ILO 2018 NRC Exam Page: 28 of 282 Question 12

Answer Explanation:

By calculation, the maximum pressures encountered for each of the conditions is as follows:

• A - 90 psi above the normal pressure regulator setpoint (944 - 949 psig) is 1034 - 1039 psig.

• B - The Reactor Pressure Scram setpoint is 1093 psig.

• C - 100 psi above the highest SRV safety setpoint (1155 psig) is 1255 psig.

• D - 175 psi above the highest SRV safety setpoint is 1330 psig.

• E - 225 psi above the highest SRV safety setpoint is 1380 psig.

(1) LCO 3.4.1.1 pressure is 1045 psig, and (2) SL 2.1.2 pressure is 1325 psig. Only condition A results in pressure <(1), and all of the other conditions are > (1). Conditions D and E would result in exceeding (2).

Distracter Explanation:

A. Answer is plausible and incorrect because the examinee could incorrectly determine that 100# above the pressure regulator setpoint would result in exceeding 1045 psig.

C. Answer is plausible and incorrect because the examinee could incorrectly determine that 100# above the pressure regulator setpoint would result in exceeding 1045 psig and/or the examinee could incorrectly determine that, since condition C resulted in exceeding 1250 psig, which is the design pressure of the RPV suction piping, that the Reactor Pressure Safety Limit was exceeded which is a common misconception. However, the SL is based on not exceeding 110% of this pressure.

D. Answer is plausible and incorrect because the examinee could incorrectly determine that, since condition C resulted in exceeding 1250 psig, which is the design pressure of the RPV suction piping, that the Reactor Pressure Safety Limit was exceeded which is a common misconception. However, the SL is based on not exceeding 110% of this pressure.

Reference Information:

S.L. 2.1 T.S. 3.4.11 NUREG 1123 KA Catalog Rev. 2 295007 High Reactor Pressure 295025 EK1. Knowledge of the operational implications of the following concepts as they apply to High Reactor Pressure:

295025 EK1.05 Exceeding safety limits G2.2.22 Knowledge of limiting conditions for operations and safety limits Technical Specifications 2.1 SAFETY LIMITS (SLs) 3.4.11 Reactor Steam Dome Pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage ILO 2015 Exam NRC Question Use (ILO 2018)

BANK High RO ILO 2018 NRC Exam Page: 29 of 282 Question 12

Associated objective(s):

Performance Objectives Cognitive Enabler Explain the operational implication as it applies to a failed safety relief valve Performance Objectives Performance Terminal Upon recognizing the entry conditions for a safety relief valve failure, operators respond to control critical parameters within expected limits IAW approved plant procedures within validated completion times ILO 2018 NRC Exam Page: 30 of 282 Question 12

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 31 of 282 Question 13

13 K/A Importance: 3.5/3.8 Points: 1.00 R13 Difficulty: 3.00 Level of Knowledge: High Source: BANK 74274 Following a transient, the following conditions exist:

• Drywell Pressure is 4.0 psig and rising.

• Drywell Temperature is 300ºF and rising.

• Torus Pressure is 3.0 psig and rising.

• Torus Temperature is 98ºF and rising.

After the LNO takes the expected manual actions for the above conditions, which one of the following reflects the expected containment response?

A. Drywell Pressure will lower due to Drywell Spray Initiation.

B. Drywell Temperature will lower due to restoring EECW to the Drywell.

C. Torus Pressure will lower due to ONLY initiating Torus Sprays.

D. Torus Pressure and Torus Temperature will lower due to operation of Torus Spray AND Torus Cooling.

Answer: D ILO 2018 NRC Exam Page: 32 of 282 Question 13

Answer Explanation:

With Torus Temperature > 95ºF, it is required to initiate Torus Cooling. Before Torus Pressure reaches 9 psig, it is required to operate Torus Sprays.

Distractor Explanation:

A is plausible; Drywell Sprays may be used in conditions of high Drywell Pressure and Temperature, Presented conditions forbid this action due to Drywell Spray Initiation Limit violation.

B is plausible; EECW may be restored to the Drywell during conditions of High Drywell Temperature.

Presented conditions forbid this action due to Drywell Temperature exceeding 242ºF.

C is plausible; Torus Spray initiation is required due to the conditions in the stem of the question which will result in a reduction in Torus Pressure. However, Torus Cooling is also required and will result in a reduction in Torus Temperature.

Reference Information:

29.100.01 SH 2 DWSIL curve to be provided.

Objective Link:

LP-OP-802-3004-0010 Plant Procedures 29.100.01 SH 2 NUREG 1123 KA Catalog Rev. 2 230000 A1. Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI: TORUS/SUPPRESSION POOL SPRAY MODE controls including:

230000 A1.02 Suppression pool temperature 295026 EK3. Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

295026 EK3.03 Suppression pool spray: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage ILO 2009 Exam NRC Question Use (ILO 2018)

BANK High RO Associated objective(s):

Primary Containment Control Performance Terminal Using the graphs from the EOP Flowcharts, and parameter values for specific plant conditions, determine appropriate operator actions per the EOP Flowcharts.

ILO 2018 NRC Exam Page: 33 of 282 Question 13

14 K/A Importance: 3.9/3.8 Points: 1.00 R14 Difficulty: 3.00 Level of Knowledge: High Source: BANK 74326 The reactor was operating at 100% power when a plant transient occurred. Several minutes later the following conditions are observed:

• Drywell pressure is 0.9 psig and rising very slowly.

• Drywell temperature is 155°F and slowly rising.

• Reactor water level is 28 inches.

Which of the following actions will be the MOST EFFECTIVE means of mitigating the rising drywell temperature trend?

A. Manually initiate both divisions of EECW.

B. Unisolate EECW to and from the drywell.

C. Manually start all single speed drywell cooling fans.

D. Increase cooling water flow using P42-F400, RBCCW Temp Control Vlv in AUTO or MANUAL.

Answer: C ILO 2018 NRC Exam Page: 34 of 282 Question 14

Answer Explanation:

Rx water is less than the Level 1 setpoint (32") and results in DW single speed fans tripping. DW fans must be restarted to lower temperature and pressure.

Distractor Explanation:

A. is incorrect because all single speed fans have tripped and cooling water would have no affect on drywell temperature/pressure without fans.

B. is incorrect because a high drywell signal does not exist and with no fans, cooling water would have no impact.

D. is incorrect because all single speed fans have tripped per Level 1 signal.

Reference Information:

29.ESP.01, page 4 29.100.01 S2 DWT leg Plant Procedures 17D41 29.ESP.01 NUREG 1123 KA Catalog Rev. 2 295028 EA1. Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE :

295028 EA1.02 3.9/3.8 Drywell ventilation system 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK High RO Associated objective(s):

EOP Objectives Cognitive Terminal Given a copy of 29.100.01 sheet 2, discuss the operator's actions for Drywell Temperature Control (DWT)

ILO 2018 NRC Exam Page: 35 of 282 Question 14

15 K/A Importance: 3.7/3.9 Points: 1.00 R15 Difficulty: 4.00 Level of Knowledge: High Source: BANK 74327 With a Torus Water Level of < -112 inches (1) which of the following Reactor Pressures require EOP action and (2) what action must be taken?

A. (1) 50 PSIG.

(2) Depressurize the RPV by opening 5 SRVs (ADS preferred)

B. (1) 50 PSIG.

(2) Rapidly depressurize the RPV ignoring cooldown rates using the Main Condenser and other steam Loads.

C. (1) 500 PSIG.

(2) Depressurize the RPV by opening 5 SRVs (ADS preferred)

D. (1) 500 PSIG.

(2) Rapidly depressurize the RPV ignoring cooldown rates using the Main Condenser and other steam Loads.

Answer: D ILO 2018 NRC Exam Page: 36 of 282 Question 15

Answer Explanation:

Torus level of -115 inches would be below the level of the SRV T-Quenchers and could possibly pressurize the Torus there per ED-7 SRV is not used and rapidly depressurize the RPV ignoring cooldown rates using another system is requried. The need to depressuirze is required per the EOPs TWL-5 Distractor Explanation:

50 PSIG is incorrect and plausible because the branch at ED-6 shows that no action is needed if already at 50 PSIG or below 50 psig is a significant number for the candidates to know because it defines the term "Decay Heat Removal Pressure (DHRP)" for Fermi 2. Per the BWROG EPGs, Appendix B, the definition of DHRP is "The lowest differential pressure between the RPV and the suppression chamber at which steam flow through the Minimum Number of SRVs Required for Emergency Depressurization (MNSRED) is sufficient to remove all decay heat from the core." The EPGs then go on to say that "the DHRP is utilized to define the depressurized state of the RPV."

The candidate should recognize that, if RPV pressure is above 50 psig then actions need to be taken to depressurize below the DHRP. If pressure is at or below the DHRP, then additional actions to depressurize are not necessary since, by definition, the plant is depressurized.

5 SRVs is plausible because this is the normal method of emergency depresurization. However per answer explanation this is incorrect.

Plant Procedures 29.100.01 SH 3 NUREG 1123 KA Catalog Rev. 2 295030 EA2. Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL :

295030 EA2.03 Reactor pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.

NRC Question Use (ILO 2018)

BANK High RO Associated objective(s):

RPV Control Performance Terminal Describe the plant condition that would require use of the alternative actions contained in 29.100.01 Sh 1, 1A, 3, and 3A, including:

a. Alternate Level Control

b. Emergency RPV Depressurization

c. RPV Flooding

d. deleted

e. Steam Cooling

f. RPV Control - ATWS ILO 2018 NRC Exam Page: 37 of 282 Question 15

16 K/A Importance: 3.9/4.6 Points: 1.00 R16 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 74366 If the following RPV Water Level actuations occurred at the level given, which one would require entry into Technical Specifications LCO 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation?

A. L1 @ 33 inches B. L2 @ 109 inches C. L3 @ 175 inches D. L8 @ 215 inches Answer: B ILO 2018 NRC Exam Page: 38 of 282 Question 16

Answer Explanation:

Per TR 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT)

Instrumentation Because the L2 trip is occuring below the setpoint at least 2 instruments must not be tripping at the required setpoint (logic is (A and C) or (B and D)). Therefore at least 2 instruments are INOPERABLE.

TS LCO 3.3.4.1 Two channels per trip system for each ATWS-RPT instrumentation Function listed below shall be OPERABLE: a. Reactor Vessel Water Level -Low Low. Level 2: and b. Reactor Vessel Pressure-High. Therefore entry is required into this LCO Distractor Explanation:

Distractors are plausible and incorrect because:

L1 is > 31.8 inches so this trip is ok, however L1 is for ECCS and ADS not ATWS-PRT L3 is > 173.4 inches so this trip is ok, however L3 is for RPS and NSSSS not ATWS-PRT L8 is < 214 inches so this trip is NOT ok, and L8 is for Main Turbine Protection.

Referece Inforamation:

TS 3.3.4.1 LCO TR 3.3.4.1 Trip Setpoint 23.601 (L1,L2,L3,L8)

Plant Procedures 23.601 NUREG 1123 KA Catalog Rev. 2 295031 Reactor Low Water Level.

G2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications Technical Specifications 3.3.4.1 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

Low NEW NRC Early Review RO ILO 2018 NRC Exam Page: 39 of 282 Question 16

Associated objective(s):

Technical Specifications for Licensed Operators Performance Enabler Given plant conditions that constitute non-compliance with any LCO, apply Technical Specifications and Bases to determine the applicable Condition(s), Required Action(s), and associated Completion Time(s).

ILO 2018 NRC Exam Page: 40 of 282 Question 16

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 41 of 282 Question 17

17 K/A Importance: 3.6/3.6 Points: 1.00 R17 Difficulty: 3.00 Level of Knowledge: High Source: NEW 74806 A reactor scram condition occurred while the plant was operating at 100% power. The Mode Switch was taken to Shutdown; however, reactor power remained above the APRM downscale trip setpoint.

The P603 Operator is attempting to determine Control Rod status from the Rod Worth Minimizer.

Under which condition, if any, would manual action be required by the P603 Operator to view the Confirm Shutdown screen on the Rod Worth Minimizer?

A. Manual action is NEVER required.

B. Manual action is ALWAYS required.

C. If BOTH RPS Trip Systems failed to actuate.

D. If BOTH RPS Trip Systems actuated but ARI failed to actuate.

Answer: C ILO 2018 NRC Exam Page: 42 of 282 Question 17

Answer Explanation:

The Confirm Shutdown display, which is part of the Rod Worth Minimizer system, provides Rod Control and Information data to the reactor operator. The screen is informational only and provides the operator with all-rods-in or not-in status with control rod I.D. and position for any not full-in control rod(s). Per 23.608, Enclosure A, the Confirm Shutdown display is automatically displayed whenever a full scram takes place.

The full scram signal for the RWM Confirm Shutdown screen comes from RPS Backup Scram logic, via relay K21A, which can be seen on I-2115-22 at grid F-8. Relay K21A is part of RPS A backup scram logic; however, it only energizes on a Full Scram signal when relays K14A or K14C de-energize AND K14F or K14H de-energize, which can be seen on I-2155-11 at grid F-7. The K14 relays de-energize upon a half scram from RPS A (K14A/K14C) or from RPS B (K14F/K14H). Both RPS trip systems, A and B, must trip (Full Scram) to energize relay K21A and actuate the Confirm Shutdown screen on the RWM automatically.

Therefore, the candidate should recall that this screen is automatically displayed ONLY when a Full Scram takes place, which the candidate should determine is when BOTH RPS Trip Systems actuate so, conversely, the candidate should determine that manual action will be necessary if BOTH RPS Trip Systems failed to actuate. Note: The Confirm Shutdown screen is available at any time to the operator through the use of soft keys on the front of the RWM even if it fails to display automatically.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly recall that the Confirm Shutdown screen actuates off of the Mode Switch position, which is plausible because of the multiple functions performed by the Mode Switch.

However, this distractor is incorrect because relay K21A actuates on a Full Scram (trip of RPS A and B) condition to display the Confirm Shutdown Screen.

B. The candidate could fail to recall that the Confirm Shutdown screen automatically displays on a Full Scram condition and therefore determine that manual action is always required to view the Confirm Shutdown screen. This is a common misconception, and is plausible, because of the large number of RPS failure scenarios in which the candidates participate while in the Simulator. These scenarios require the candidate to manually actuate the Confirm Shutdown screen using the soft keys on the front of the RWM. The candidate could recall the numerous times that he/she had to manually actuate the Confirm Shutdown screen and conclude that it must always be done, which is incorrect since the Confirm Shutdown screen does automatically actuate on a Full Scram condition by the relay logic described above.

D. The candidate could incorrectly recall that the Confirm Shutdown screen actuates off of ARI Logic instead of RPS Logic, which is plausible because of the close interrelationships that exist between ARI and RPS. However, this distractor is incorrect because relay K21A actuates on a Full Scram (trip of RPS A and B) condition to display the Confirm Shutdown Screen and not from ARI logic.

Reference Information:

23.608, Rod Worth Minimizer System SOP, Enclosure A.

I-2115-22, RWM Schematic Diagram.

I-2155-11, RPS Scram Disch Vol Isol Valve Ind Lights, Backup Scram Valves.

ILO 2018 NRC Exam Page: 43 of 282 Question 17

NUREG 1123 KA Catalog Rev. 2 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or unknown.

295037 EA1. Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :

295037 EA1.08 3.6/3.6 Rod control and information system: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Rod Worth Minimizer (C1108)

Cognitive Enabler Discuss effective monitoring of the Rod Worth Minimizer system using local, remote, computer displays and alarms.

ILO 2018 NRC Exam Page: 44 of 282 Question 17

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 45 of 282 Question 18

18 K/A Importance: 3.3/4.0 Points: 1.00 R18 Difficulty: 3.00 Level of Knowledge: Low Source: BANK 74526 Why does the Radioactivity Release Control Leg of the EOPs permit the RESTART of isolated HVAC Systems?

While executing the Radioactivity Release Control Leg, restarting HVAC Systems:

A. ensures a positive pressure is maintained in the Control Room.

B. ensures accessibility is maintained inside the Secondary Containment.

C. provides filtration and adsorption of radioactivity via an elevated release path.

D. ensures accessibility is maintained in buildings outside the Secondary Containment.

Answer: D ILO 2018 NRC Exam Page: 46 of 282 Question 18

Answer Explanation:

Continued personnel access to the turbine building, auxiliary building, or other buildings outside the secondary containment boundary may be essential for responding to emergencies or transients which may degrade into emergencies. These buildings are not always airtight structures, and radioactivity release inside the buildings would not only limit personnel access, but would eventually lead to an unmonitored ground level release. Operating HVAC preserves building accessibility and discharges radioactivity through an elevated, monitored release point.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. the Control Center is outside of Secondary Containment and limiting radiation levels inside the Control Center Envelope is a primary concern during radiation release events. However, 29.ESP.25, Defeat of Radwaste and TBHVAC Radiation Isolations, does NOT include Control Center HVAC.

B. safety related equipment is located INSIDE the Secondary Containment and acting to preserve accessibility to this equipment is important during radiation release events. The candidate could confuse the action in the Secondary Containment Control EOP that permits restarting RBHVAC to preserve access to safety related equipment, however, that is not the basis behind the actions taken in the Radioactivity Release Control EOP.

C. this is the reason why the Standby Gas Treatment System is started while executing the Secondary Containment Control EOP, so the candidate could incorrectly determine that this is the basis for the action taken in the Rad Release EOP. However, although starting HVAC outside of secondary containment does provide an elevated release path, these HVAC systems (RWHVAC and TBHVAC) do NOT provide for filtration and/or adsorption of radioactivity.

Reference Information:

ST-OP-802-3005-001, Secondary Containment Control and Rad Release Student Text.

BWROG EPGs/SAGs, Appendix B, Page B-9-3.

29.ESP.25, Defeat of Radwaste and TBHVAC High Radiation Isolations.

NUREG 1123 KA Catalog Rev. 2 295038 High Off-Site Release Rate G2.4.18 Knowledge of the specific bases for EOPs 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

BANK Low RO Associated objective(s):

Secondary Containment Control and Radioactive Release Performance Terminal State the basis for the Override Statements contained in 29.100.01 Sh 5, Secondary Containment Control and Radioactive Release.

ILO 2018 NRC Exam Page: 47 of 282 Question 18

19 K/A Importance: 2.8/3.4 Points: 1.00 R19 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 74446 A plant fire is in progress. Operators are currently implementing AOP 20.000.22, Plant Fires.

The CRS then orders the Action of Condition C to PERFORM 20.000.18, Control of the Plant from the Dedicated Shutdown Panel.

What is the reason for performing this action?

A. The operators need to evacuate the Main Control Room because the fire has caused it to become uninhabitable.

B. The need to shut down the plant from the Dedicated Shutdown Panel exists due to personnel feeling the effects of smoke inhalation.

C. There is the possibility of equipment malfunctions and safe operation of the plant may be jeopardized due to a fire in one or more of the 3L zones.

D. The possibility of losing adequate core cooling exists because High Pressure Feed sources are insufficient to maintain RPV Level 3 and two or more SRVs are lifting.

Answer: C ILO 2018 NRC Exam Page: 48 of 282 Question 19

Answer Explanation:

This question requires the candidate to know the reason why Action C of 20.000.22, Plant Fires is performed.

The reason that Action C of 20.000.22, Plant Fires is performed is stated in the 20.000.22 Bases document section for Action C, which states "The alternative shutdown system was designed to provide safe-shutdown capability separate and remote from the control center complex, when a fire in the complex or the other 3L zones is assumed to significantly damage the equipment or cabling in these zones. If there is a fire in one or more of the 3L zones, then there is the possibility of equipment malfunctions and safe operation of the plant may be jeopardized. It is for this reason that 20.000.18 is entered."

Therefore, the operator should determine that this action has been directed by the CRS because of the possibility of equipment malfunctions and safe operation of the plant may be jeopardized when a fire in a 3L zone of concern is detected.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could confuse the basis for performing Condition C with the basis for entering 20.000.19, Shutdown from Outside the Main Control Room. This could lead the candidate to conclude that, if in the Fire AOP and if the MCR becomes uninhabitable, that control would be shifted to the Dedicated Shutdown Panel. This is incorrect because control would be shifted to the Remote Shutdown Panel, NOT the Dedicated Shutdown Panel, and procedure 20.000.19 performed rather than 20.000.18 as stated in the stem of the question.

B. The candidate could recognize the condition in this distractor as being a condition (IF statement) for one of the Overrides in 20.000.22 and conclude that the correct action to take, if personnel in the MCR are feeling the effects of smoke inhalation, is to transition control to the Dedicated Shutdown Panel. This is incorrect because the Override requires all personnel to don respirators under this condition.

D. The operator could recognize the condition in this distractor as being a condition (IF statement) for one of the Overrides in 20.000.22. This could lead the candidate to conclude that the correct action to take, if the possibility of losing adequate core cooling due to a loss of adequate HP Feed sources with two open SRVs exists, is to transition control to the Dedicated Shutdown Panel. This is incorrect because the Override requires the operators to enter the EOPs for RPV Level Control.

Reference Information:

20.000.22, Plant Fires AOP 20.000.22 Bases - Plant Fires AOP BASES 20.000.19, Shutdown from Outside the Main Control Room 20.000.19, Shutdown from Outside the Main Control Room BASES NUREG 1123 KA Catalog Rev. 2 600000 Plant Fire On Site 600000 AK3. Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE:

600000 AK3.04 Actions contained in the abnormal procedure for plant fire on site 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

Low NEW RO ILO 2018 NRC Exam Page: 49 of 282 Question 19

Associated objective(s):

Reactor Operator Exam Objectives Performance Enabler Explain bases for notes, cautions and overrides.

ILO 2018 NRC Exam Page: 50 of 282 Question 19

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 51 of 282 Question 20

20 K/A Importance: 3.3/3.4 Points: 1.00 R20 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 74406 Following a Grid Disturbance and plant transient, Main Turbine Generator (MTG) conditions are as shown:

What is the current operating condition of the MTG and what action could the operator take to restore MTG MVARs to the GREEN BAND?

A. The MTG is currently over-excited; Operate the Turbine Speed/Load Controls in the raise direction.

B. The MTG is currently over-excited; Operate the Voltage Reg Control SW 90CS in the raise direction.

C. The MTG is currently under-excited; Operate the Turbine Speed/Load Controls in the raise direction.

D. The MTG is currently under-excited; Operate the Voltage Reg Control SW 90CS in the raise direction.

Answer: D ILO 2018 NRC Exam Page: 52 of 282 Question 20

Answer Explanation:

23.118 Enclosure G indicates that the conditions shown in the stem of the question place the MTG in an Under-excited (Leading Power Factor) condition.

To cause MVARS out of the machine, and restore MTG MVARs to the GREEN BAND, the operator would raise the Voltage Reg Control SW 90CS until VARS are positive (out). This guidance can be seen on Page 27 of 23.118.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could determine that the conditions shown place the MTG in an over-excited condition, which is incorrect as can be seen on 23.118, Enclosure G. The candidate could also conclude that the Turbine Speed/Load Controls must be placed in the raise direction to restore MTG MVARs to the green band, which is incorrect as described in 23.118.

B. The candidate could determine that the conditions shown place the MTG in an over-excited condition, which is incorrect as can be seen on 23.118, Enclosure G. The candidate could correctly conclude that the Voltage Reg Control Switch must be placed in the raise direction to restore MTG MVARs in the green band.

C. The candidate could determine that the conditions shown place the MTG in an under-excited condition, which is correct as can be seen on 23.118, Enclosure G. The candidate could then conclude that the Turbine Speed/Load Controls must be placed in the raise direction to restore MTG MVARs to the green band, which is incorrect as described in 23.118.

Reference Information:

23.118, Main Generator and Generator Excitation System SOP, Enclosure G (Generator Operating Guidelines).

NUREG 1123 KA Catalog Rev. 2 700000 Generator Voltage and Electric Grid Disturbances 700000 AK1. Knowledge of the operational implications of the following concepts as they apply to GENERATOR VOLTAGE AND ELECTRICAL GRID DISTURBANCES:

700000 AK1.03 3.3/3.4 Under-excitation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (4) Secondary coolant and auxiliary systems that affect the facility.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Main Generator and Excitation (N3031 & N3034)

Cognitive Enabler Identify associated remote and local instrumentation, indications, alarms, and controls for the Main Generator and Excitation System.

Main Generator and Excitation (N3031 & N3034)

Cognitive Enabler Identify associated remote and local instrumentation, indications, alarms, and controls for the Main Generator and Excitation System.

ILO 2018 NRC Exam Page: 53 of 282 Question 20

21 K/A Importance: 4.5/4.7 Points: 1.00 R21 Difficulty: 3.00 Level of Knowledge: Low Source: NEW 74386 Which of the following is a Symptom for entry into Abnormal Operating Procedure 20.125.01, Loss of Condenser Vacuum?

A. Decreasing Offgas Outlet Flow.

B. Decreasing Offgas Delay Pipe Pressure.

C. Decreasing Main Generator Megawatts..

D. N62-F406A, North Ringwater Vac Pump Recirc Line PCV indicates full OPEN.

Answer: C ILO 2018 NRC Exam Page: 54 of 282 Question 21

Answer Explanation:

20.125.01, Loss of Condenser Vacuum, lists the following as symptoms that require entry into this AOP:

• Degrading condenser vacuum as indicated on N30-R824, Main Cond Vacuum Pressure Recorder and/or N30-R823, Main Cond Vacuum Pressure Indicator

• Main Generator MW decreasing

• 4D108, CONDENSER PRESSURE HIGH

• 6D16, OFF GAS SYS MN CONDENSER PRESSURE HIGH The candidate should recognize that, even though each of the Distractors are indications and/or components that would respond to degrading Off Gas and condenser vacuum, decreasing MTG MW is the only one that is explicitly listed as symptom.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Off Gas flow changes in response to Condenser air in-leakage and degradations in the Off Gas system can lead to a loss of condenser vacuum. The candidate could determine that the decreasing Off Gas flow is either the result of, or could be the cause of, degrading condenser vacuum and therefore conclude that AOP entry is required. However, this indication is not a symptom listed in 20.125.01 requiring entry into the AOP.

B. Off Gas Delay Pipe Pressure changes in response to Condenser air in-leakage and degradations in the Off Gas system can lead to a loss of condenser vacuum. The candidate could determine that the decreasing Delay Pipe Pressure is either the result of, or could be the cause of, degrading condenser vacuum and therefore conclude that AOP entry is required. However, this indication is not a symptom listed in 20.125.01 requiring entry into the AOP.

D. The N62-F406A changes position in response to changes in Off Gas Delay Pipe pressure. The candidate could assume that the F046A opens more to increase flow through Off Gas in response to increasing Condenser air in-leakage and, therefore, associate this indication as being a symptom of AOP 20.125.01; however, the N62-F406A going open is not a symptom listed in 20.125.01.

Reference Information:

20.125.01, Loss of Condenser Vacuum.

NUREG 1123 KA Catalog Rev. 2 295002 Loss of Main Condenser Vacuum G2.4.4 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Reactor Operator Performance Enabler Appropriately recognize symptoms for entry into emergency and abnormal operating procedures.

ILO 2018 NRC Exam Page: 55 of 282 Question 21

22 K/A Importance: 3.1/3.4 Points: 1.00 R22 Difficulty: 3.00 Level of Knowledge: High Source: NEW 74306 The plant has been operating at 100% power for 300 consecutive days, when the following events occur:

• A reactor scram causes the crew to place the Mode Switch in Shutdown.

• The MSIVs closed 10 minutes later.

• RPV Pressure is currently 1095 psig.

• The CRLNO has been directed to maintain RPV Pressure 900-1050 psig.

How many SRVs are going to be required to restore and maintain RPV pressure in the desired control band considering heat input by DECAY HEAT GENERATION ONLY?

A. One SRV opening and closing at it's Low-Low Set setpoints.

B. One open SRV will restore RPV pressure in band, then it can be closed for the duration of the event.

C. One Low-Low Set SRV will be open continuously, with the second opening and closing at it's Low-Low Set setpoints.

D. Both Low-Low Set SRVs will be open continuously with the CRLNO opening and closing one additional SRV.

Answer: A ILO 2018 NRC Exam Page: 56 of 282 Question 22

Answer Explanation:

The conditions in the stem of the question have caused a High Reactor Pressure. The candidate will have to determine the impact of decay heat generation so as to correct the High Reactor Pressure and control RPV Pressure in the designated control band for the duration of the event.

Decay heat produced is at a level dependent on power history. From a scram at 100% power, initially the thermal output of the reactor will be about 100% power, 7% supplied by decay heat. Thermal heat output decreases rapidly to the decay heat level. Eight to ten seconds after the scram, thermal output is due mainly to decay heat and drops to 7% of rated thermal output. After approximately one minute, thermal output is 3 to 5% of rated and drops to about 2% after about ten minutes. One hour after a scram, decay heat is about 1% rated thermal output. The information given in the stem of the question indicates that the reactor has been shut down for about 10 minutes, so decay heat generation will be between approximately 2%.

SRV capacity (steam flow) is given in UFSAR Paragraph 5.2.2.3.3.4 as 87E4 lb/hr (or 870,000 lb/hr) at 1090 psig, which is approximately where the stem of the question has the plant. UFSAR Section 1.2.2, Plant Description, shows that rated steam flow at Fermi 2 is 14.9 lb/hr at 991 psia. 2% of this value is 298,000 lb/hr, which is within the capacity of one SRV. Therefore, the candidate should determine that one SRV opening and closing at its Low-Low setpoint will be enough to control the steam generated from decay heat approximately 10 minutes after plant shutdown from 100% power operation.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could determine that there is no significant decay heat generation 10 minutes after shutdown and, once RPV pressure is lowered back in band by opening an SRV, it will remain in band with no further SRV actuations. This is incorrect because decay heat, in the 2% range, exists and will continue to exist at the 1% value one hour after shutdown, so further SRV actuation must occur to control RPV pressure.

C. The candidate could determine that the decay heat generated 10 minutes after shutdown is above the capacity of one, but within the capacity of two Low-Low Set SRVs. This is a common misconception because most candidates readily remember that decay heat drops to 7% within several seconds of a plant shutdown, due to the decay of delayed neutrons and they determine that 7% is the capacity of about one and a half SRVs. However, as stated above, after approximately ten minutes, thermal output due to decay heat drops to about 2%, which is within the capacity of one Low-Low Set SRV.

D. The candidate could determine that the decay heat generated 10 minutes after shutdown is above the capacity of both Low-Low Set SRVs, which would require the operator to manually actuate one additional SRV to control RPV pressure in band. This is incorrect, however, as stated above, because after approximately ten minutes, thermal output due to decay heat drops to about 2%, which is within the capacity of one Low-Low Set SRV.

Reference Information:

BR08Sr5_Operational_Physics May 2011 - Reactor Operational Physics GFE Student Text, Page 33 description of decay heat generation after shutdown from 100% power.

UFSAR Section 1.2.2, Plant Description.

UFSAR Section 5.2.2.3.3.4, Safety/Relief Valve Characteristics.

ILO 2018 NRC Exam Page: 57 of 282 Question 22

NUREG 1123 KA Catalog Rev. 2 295007 High Reactor Pressure 295007 AK1. Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR PRESSURE:

295007 AK1.02 3.1/3.4 Decay heat generation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High NEW RO ILO 2018 NRC Exam Page: 58 of 282 Question 22

23 K/A Importance: 3.6/3.7 Points: 1.00 R23 Difficulty: 4.00 Level of Knowledge: High Source: NEW 74346 The Mode Switch has been taken to shutdown at 30% power for the start of RF-19. The following plant conditions currently exist:

• Post Scram Feedwater Logic actuated on the Scram.

• Reactor Pressure has been lowered to 500 psig.

• 3D148, FW/MTG RPV H2O Level 8 Trip is in alarm.

• RPV water level is 216" and slowly rising.

• 20.000.23, High RPV Water Level has been entered.

Which of the following Reactor Water Level Control actions should be taken, in accordance with the High RPV Water Level AOP, to mitigate these conditions?

A. Lower SETPOINT on C32-R618, Master Feedwater Level Controller.

B. Lower DEMAND on C32-R620, N21-F403 RPV Startup LCV Controller.

C. Lower DEMAND on C32-R616A (B), N (S) Reactor Feed Pump Controllers.

D. Lower SETPOINT on N21-K858A (B), N (S) Reactor Feed Pump Min Flow Controllers.

Answer: B ILO 2018 NRC Exam Page: 60 of 282 Question 23

Answer Explanation:

The conditions given in the stem of the question indicate that a high RPV water level condition has occurred. AOP 20.000.23, High RPV Water Level, provides actions that can be taken, for a high RPV water level condition, when leakage past the Startup Level Control Valve (SULCV) is suspected. For this condition, the operator should attempt to lower the demand on the SULCV Controller to minimum (-5) to attempt to fully seat the valve. The candidate should recognize that this action should be taken for the conditions given in the stem of the question.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The C32-R618, Master Feedwater Level Controller, is the controller that is normally manipulated to control components associated with the Feedwater Control System, and the candidate could assume that lowering the setpoint of this controller would have the desired impact. However, when a scram occurs, Post Scram Feedwater Logic actuates as described in ARP 3D157, Post Scram FW Logic Actuated. One of the results of this is that the C32-R616A and B (N and S RFP Controllers) get set to MANUAL mode. When this occurs, the C32-R618 also gets set to MANUAL, so it no longer controls the RFP speeds. Also, as stated in Section 1.1 of 23.107, this controller can NOT be placed in AUTO until one of the RFP controllers is placed in AUTO, so lowering its setpoint will not have any impact.

C. After a scram, the C32-R616A and B (N and S RFP Controllers) get set to MANUAL mode. When this occurs, the operator can adjust the output of these controllers to change RFP speed. The candidate could determine that lowering the output of these controllers will have the desired impact. However, the stem of the question states that RPV level is 216", which is above the High RPV Water Level Trip Setpoint of 214" (Level 8), so the RFPs will be tripped and lowering their controller outputs will have no impact.

D. The candidate could determine that leakage is occurring through the RFP Min Flow lines, which could be impacting RPV Water Level. The candidate could assume that lowering the Setpoint of these controllers will have the desired impact. However, the stem of the question states that RPV level is 220", which is above the High RPV Water Level Trip Setpoint of 214" (Level 8), so the RFPs will be tripped. When an RFP trips, its Minimum Flow Line motor operated valve (MOV) goes shut to isolate the minimum flow line, so adjusting these controllers will have no impact.

Reference Information:

20.000.32, High RPV Water Level AOP.

23.107, Reactor Feedwater and Condensate Systems, Section 1.1, System Description for the Feedwater Control System.

ARP 3D157, Post Scram FW Logic Actuated.

NUREG 1123 KA Catalog Rev. 2 295008 High Reactor Water Level 295008 AK2. Knowledge of the interrelations between HIGH REACTOR WATER LEVEL and the following:

295008 AK2.03 3.6/3.7 Reactor water level control 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Feedwater Control (C3200)

Cognitive Enabler Discuss failure modes of Feedwater Control System controls and vital instruments, including design features that could result in erroneous operation or indication.

ILO 2018 NRC Exam Page: 61 of 282 Question 23

24 K/A Importance: 3.9/4.2 Points: 1.00 R24V2 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 75128 Which of the following Process Radiation Monitors, if verified to be at their high-high or upscale trip setpoints, would require entry into 29.100.01, Sheet 5, Secondary Containment Control on Secondary Containment Exhaust High Radiation?

A. D11-K601A (B) Channel A (B) Off Gas Radiation Monitors.

B. D11-K808 (K810) Div 1 (2) RB Vent Exhaust Radiation Monitors.

C. D11-K809 (K813) Div 1 (2) CCHVAC Makeup Air Radiation Monitors.

D. D11-K603A (B, C, D) Channel A (B, C, D) Main Steam Line Radiation Monitors.

Answer: B ILO 2018 NRC Exam Page: 62 of 282 Question 24

Answer Explanation:

The D11-K808 (K810) Div 1 (2) RB Vent Exhaust Radiation Monitors have an Upscale Trip setpoint of 16,000 cpm. If the radiation monitor is at it's trip setpoint, as stated in the stem of the question, then entry into 29.100.01 Sheet 5, Secondary Containment Control EOP is required. This requires the candidate to know the interrelationship between this Process Radiation Monitor and Secondary Containment High Radiation.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The D11-K601A (B) Channel A (B) Off Gas Radiation Monitors have upscale trip setpoints of 1,000 mr/hr. Per ARP 3D12, if this radiation monitor is at it's trip setpoint, entry into 20.000.07, Fuel Cladding Failure, is required. This distractor is plausible because the candidate could determine that, since a fuel clad failure is evidenced by this alarm, and radiation levels in secondary containment could increase as a result of the fuel failure, EOP entry is required, which is incorrect.

C. The D11-K809 (K813) Div 1 (2) CCHVAC Makeup Air Radiation Monitors have upscale trip setpoints of 150 (350) cpm respectively. Per ARP 3D45, if these radiation monitors are at their trip setpoint, entry into 20.000.02, Abnormal Release of Radioactive Material, is required. This distractor is plausible because the candidate could determine that, since an abnormal release of radioactive material has occurred, then EOP entry must also be required, which is incorrect.

D. The D11-K603A (B, C, D) Channel A (B, C, D) Main Steam Line Radiation Monitors have upscale trip setpoints of 3,100 (3,300, 3,400, 3,000) mr/hr respectively. Per ARP 3D82, if these radation monitors are at their alarm setpoints, entry into AOPs 20.000.07, Fuel Clad Failure and 20.000.21, Reactor Scram is required. The candidate could determine that the radiation level sensed by these monitors, if reached, would warrant EOP entry. Or, the candidate could determine that, since a fuel clad failure is evidenced by this trip, and radiation levels in secondary containment could increase as a result of the fuel failure, that EOP entry is required, which is incorrect.

Reference Information:

3D12 Offgas Rad Monitor Rad High-High 3D36 RB Vent Exh Rad Monitor Upscale Trip 3D45 CCHVAC Makeup Air Rad Monitor Upscale Trip 3D82 Mn Steam Line Radiation Upscale/Inop Channel Trip NUREG 1123 KA Catalog Rev. 2 295034 Secondary Containment Ventilation High Radiation 295034 EK2. Knowledge of the interrelations between SECONDARY CONTAINMENT VENTILATION HIGH RADIATION and the following:

295034 EK2.01 3.9/4.2 Process radiation monitoring system 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(11) Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Reactor Operator Performance Enabler Appropriately recognize symptoms for entry into emergency and abnormal operating procedures.

Process Radiation Monitoring System (D1100)

Cognitive Enabler Identify abnormal and emergency operating procedures associated with the Process Radiation Monitoring System.

ILO 2018 NRC Exam Page: 63 of 282 Question 24

25 K/A Importance: 3.3/3.5 Points: 1.00 R25 Difficulty: 3.00 Level of Knowledge: Low Source: NEW 74286 The control room receives annunciators 8D46 and 17D46, Div I and II Reactor Building Pressure High/Low, during a severe thunderstorm and high winds in the area.

The CRLNO observes that both the T41-R800A and B, Div 1 and 2 CR and RB Differential Pressure Recorders indicate RB differential pressure is reading +0.5 inches WC.

Which of the following indicates the response of secondary containment ventilation to these conditions as well as the reason for this response?

A. The RBHVAC Supply and Exhaust Fans will automatically trip to prevent damage to RBHVAC ductwork.

B. The RB Supply Fan blades will modulate to restore a negative differential pressure to prevent the unmonitored release of potentially contaminated air from Secondary Containment.

C. The RB Exhaust Fan inlet dampers will modulate to restore a negative differential pressure to prevent the unmonitored release of potentially contaminated air from Secondary Containment.

D. The RBHVAC Supply and Exhaust Fans will automatically trip and SGTS will automatically start to prevent damage to RBHVAC ductwork AND to restore a negative differential pressure to prevent the unmonitored release of potentially contaminated air from Secondary Containment.

Answer: C ILO 2018 NRC Exam Page: 64 of 282 Question 25

Answer Explanation:

The Secondary Containment Atmosphere Pressure Monitoring System (which is part of the RBHVAC System) modulates the inlet dampers to the exhaust fans to maintain a differential pressure in the Reactor Building with respect to the outside air of -1/4" H2O.

With high wind conditions, RB pressure can increase and cause high RB pressure alarms to come in.

The system will modulate the position of the exhaust fan dampers to restore negative RB pressure to prevent an unmonitored release of potentially contaminated air from Secondary Containment.

Distractor Explanation:

A. Is plausible because the RBHVAC Supply and Exhaust fans do trip at high RB pressure, but not until 2.5 inches WC. For a pressure this high, it is plausible to assume the fans would trip to protect the integrity of the RBHVAC ductwork.

B. Is plausible because some fans for various ventilation systems at Fermi do have, or had in the past, variable blading to control air flow. Also, the CCHVAC Compressors have variable inlet vanes to control compressor capacity. Additionally, some ventilation systems modulate the inlet dampers to their Supply Fans to maintain building pressure, but not RBHVAC. The reason for restoring RB pressure in this response is correct. Note: 2 (of 4) validators selected this response during validation, which lends further plausibility to this distractor.

D. Is plausible because the RBHVAC Supply and Exhaust fans do trip at high RB pressure, but not until 2.5 inches WC. Additionally, most RBHVAC system trips occur due to Secondary Containment Isolations, which also result in an automatic start of SGTS. The resons behind this response are plausible because it is desirable to restore negative building pressure to prevent an unmonitored release from Secondary Containment and the candidate could assume that the high RB pressure could damage the RBHVAC ductwork.

Reference Information:

23.426 RBHVAC System SOP, Section 1.1 System Description for explanation of why negative pressure is maintained in the RB as well as a list of RBHVAC fan trips.

T41-00 RBHVAC System Design Basis Document, Section 4.1.8.1.

NUREG 1123 KA Catalog Rev. 2 295035 Secondary Containment High Differential Pressure 295035 EK3. Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:

295035 EK3.02 3.3/3.5 Secondary containment ventilation response 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (9) Shielding, isolation, and containment design features, including access limitations.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Reactor Building HVAC Cognitive Enabler Describe general Reactor Building HVAC System operation, including component operating sequence, normal operating parameters, and expected system response.

Reactor Building HVAC Cognitive Enabler Describe general Reactor Building HVAC System operation, including component operating sequence, normal operating parameters, and expected system response.

ILO 2018 NRC Exam Page: 65 of 282 Question 25

26 K/A Importance: 3.5/3.6 Points: 1.00 R26 Difficulty: 3.00 Level of Knowledge: High Source: NEW 74272 The plant is operating at 97% power when a seismic event occurs. The following alarms and indications are received:

• 6D69, SEISMIC SYSTEM EVENT/TROUBLE

• 2D78, REAC BLDG FLR/EQUIP DRN SUMPS LVL HI-HI/LO-LO

• 2D105, REAC BLDG CORNER ROOMS/HPCI ROOM FLOOD LEVEL The G11-R651, SE Corner Sump G1101-D074 Level Recorder, indicates the following:

• Sump Level is steady at 31.9 inches.

• Flood Level is downscale at 5.0 inches.

The G11-R654, SW Corner Sump G1101-D076 Level Recorder, indicates the following:

• Sump Level is lowering at 44.0 inches.

• Flood Level is downscale at 5.0 inches.

The G11-R655, HPCI Room Level Recorder, indicates 39.0 inches and rising.

Both G1101-D076 RB SW Floor Drn Sump Pumps are running.

Which one of the following actions is required?

A. Close the E4150-F004, HPCI CST Suction Isolation Valve.

B. Close the E2150-F036B, Div 2 Core Spray Pumps Torus Suction Valve.

C. Close the E1150-F004B (D) RHR Pump B (D) Torus Suction Isolation Valves.

D. Close the E4150-F002 (F600), HPCI Steam Supply Inboard (Outboard Bypass)

Isolation Valves.

Answer: A ILO 2018 NRC Exam Page: 66 of 282 Question 26

Answer Explanation:

Indications are given that a leak is coming from the suction (water) side of the HPCI System. 29.100.01 Sheet 5 directs the operator to isolate systems leaking into secondary containment when the Max Norm Sump Level or Area Water Level is reached. The candidate should recall that 45 inches in the D076 sump is the Max Norm Operating Water Level in that sump and 8 inches in the HPCI Room is the Max Norm Operating Water Level, therefore conditions dictate that a system must be isolated.

Next, the candidate should determine which system to isolate. The candidate should recall that the HPCI room does not have its own floor drain sump and instead drains water, via a trough and pipe, to the D076 (SW floor drain) sump. Since this sump is located in the same room as the Div 2 RHR pumps, the candidate must determine that, since Flood Level in the SW corner is steady and flood level in the HPCI room is rising, that the leak is from the HPCI system and not Div 2 RHR. Lastly, the candidate must decide between the HPCI system and Div 2 Core Spray. Since HPCI and Div 2 Core Spray are both in the SE corner of the Reactor Building (separated by a normally closed watertight door), the candidate should recognize that the D074, SE Sump and Flood Level indicators are steady, which indicates that the leak is NOT from Div 2 Core Spray.

Note: Flood Control Valve T4500-F601 exists in the line between the HPCI room trough and the SW quadrant sump (DO76) to prevent cross-flooding of the SW Quadrant and the HPCI Room. Since the HPCI Room does not have its own dedicated sump, its floor and trench drains are directed to G1101-D076 (i.e., the SW Quad).

T4500-F601 is normally open to allow the HPCI drains to flow into D076. In the event of excessive equipment leakage or a line break, the valve will close on a D076 High/High alarm (+45"). This will prevent water from backing up into the drains and flooding the unaffected area.

Therefore, the candidate must conclude that the leak is from the HPCI water side, which must be isolated.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could recall that the HPCI system and Division 2 Core Spray are both in the Southeast corner of the Reactor Building and assume that the rising sump level and floor level are due to a leak from Div 2 Core Spray. This is incorrect because HPCI and Div 2 Core Spray are separated by a normally closed watertight door. Even though this door could be open, the stem of the question indicates an increasing HPCI room and D076 sump level and NOT an increasing D074 sump level, which is the sump that would indicate a leak from Div 2 Core Spray.

C. The candidate could fail to recall that the HPCI room does not have its own Floor Drain Sump and instead drains to the SW (D076 Sump), which is the same area as Division 2 RHR, in the Southwest (SW) corner of the Reactor Building. This could lead the candidate to focus in only on the rising D076 (SW corner room) sump level and assume that it is due to a leak from Div 2 RHR. This is incorrect because the stem of the question indicates an increasing HPCI room and D076 sump level and NOT a SW Corner Room Flood Level (measured from the floor) that would indicate a leak from Div 2 RHR.

Additionally, the stem indicates that D076 level is lowering, which the candidate should determine is due to the T4500-F601 going closed on high/high sump level (+45").

D. The candidate could determine that the leak is from the HPCI system and conclude that the steam supply isolation valves must be closed. This is incorrect because a leak from the HPCI steam supply, although it would cause an increase in sump activity (due to condensation of the steam on the walls and floor of the HPCI room), would not cause the changes in sump volume represented by the information given in the stem of the question.

Reference Information:

29.100.01 SH 5, Secondary Containment Control.

2D78, REAC BLDG FLR/EQUIP DRN SUMPS LVL HI-HI/LO-LO 2D105, REAC BLDG CORNER ROOMS/HPCI ROOM FLOOD LEVEL ILO 2018 NRC Exam Page: 67 of 282 Question 26

Plant Procedures 29.100.01 SH 5 29.ESP.01 NUREG 1123 KA Catalog Rev. 2 295036 Secondary Containment High Sump / Area Water Level 295036 EA1. Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL:

295036 EA1.02 3.5/3.6 Affected systems so as to isolate damaged portions 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Secondary Containment Control and Radioactive Release Performance Terminal State the criteria for isolating a system that is discharging into a flooded area of the Secondary Containment.

Secondary Containment Control and Radioactive Release Performance Terminal State the actions required if all the Reactor Building corner rooms were flooded to 5 feet from primary or secondary systems.

ILO 2018 NRC Exam Page: 68 of 282 Question 26

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 69 of 282 Question 27

27 K/A Importance: 3.3/3.3 Points: 1.00 R27 Difficulty: 4.00 Level of Knowledge: High Source: NEW 74906 A Design Basis Accident (DBA) Loss of Coolant Accident (LOCA) is in progress and fuel damage has occurred.

The Primary Containment Monitoring System is aligned in its NORMAL configuration.

The following primary conditions exist:

• Alarms 8D55 and 17D8, Div I and Div II Hydrogen Concentration High, are both in alarm.

• Alarms 8D54 and 17D7, Div I and Div II Hydrogen Concentration High-High, are both in alarm.

• Hydrogen Concentration in both the Drywell and Supression Chamber is 6.0% and rising.

Which of the indications below would represent a combustible mixture in the SUPPRESSION CHAMBER?

A. 8D50, Div I Oxygen Concentration High, with the T50-R806A, Div 1 Drywell Atm Analysis Recorder at 3.5% Oxygen.

B. 17D12, Div II Oxygen Concentration High, with the T50-R806B, Div 2 Drywell Atm Analysis Recorder at 3.5% Oxygen.

C. 8D51, Div I Oxygen Concentration High-High, with the T50-R806A, Div 1 Drywell Atm Analysis Recorder at 4.5% Oxygen.

D. 17D4, Div II Oxygen Concentration High-High, with the T50-R806B, Div 2 Drywell Atm Analysis Recorder at 4.5% Oxygen.

Answer: D ILO 2018 NRC Exam Page: 70 of 282 Question 27

Answer Explanation:

From the basis for TS LCO 3.6.3.1, Primary Containment Oxygen Concentration:

"The primary containment (both drywell and suppression chamber) oxygen concentration is maintained <

4.0% to ensure that an event that produces any amount of hydrogen does not result in a combustible mixture inside primary containment."

Therefore, as long as Primary Containment Oxygen Concentration remains below 4.0%, a combustible mixture is not present.

The stem of the question gives that Hydrogren Concentration is above 4%, which is the MEL (minimum explosive limit) for Hydrogen. The candidate must recognize the significance of Oxygen concentration being above 4% in the Suppression Chamber with >4% Hydrogen.

Division I PCMS is normally alligned to monitor the Drywell atmosphere. Division II PCMS is normally aligned to monitor the Suppression Chamber, therefore only the Division II indications (alarms and indications on the H11-P817) would indicate Oxygen above 4% in the Suppression Chamber.

The setpoint for 17D12, Div II Oxygen Concentration High is 3.5%. With Oxygen at the alarm setpoint, a combustible mixture is NOT present.

The setpoint for 17D4, Div II Oxygen Concentration High-High is 4.5%. With Oxygen at the alarm setpoint, a combustible mixture IS present.

Therefore, the candidate must recognize that a combustible mixture is only present, given the High Primary Containment Hydrogen Concentration in the stem of the question, when a valid 17D12 alarm is received, as evidenced by 4.5% Oxygen Concentration.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly determine that 3.5% Oxygen Concentration would form a combustible mixture with 6% Hydrogen. However, this is not true with only 3.5% Oxygen. Also, these are Division I indications, which are indicative of Drywell Oxygen and not the Suppression Chamber.

B. These are Division II indications, so they are indicative of Suppression Chamber Oxygen Concentration. The candidate could incorrectly determine that 3.5% Oxygen Concentration would form a combustible mixture with 6% Hydrogen. However, this is not true with only 3.5% Oxygen.

C. The indications given are above 4% Oxygen and would represent a combustible mixture given 6%

Hydrogen. However, these are Division I indications, which are indicative of Drywell Oxygen Concentration and not the Suppression Chamber.

Reference Information:

Tech Spec Bases 3.6.3.1, Primary Containment Oxygen Concentration.

ARPs 8D55 and 17D8, Div I and Div II Hydrogen Concentration High.

ARPs 8D54 and 17D7, Div I and Div II Hydrogen Concentration High.

ARP 8D50, Div I Oxygen Concentration High.

ARP 8D51, Div I Oxygen Concentration High-High.

ARP 17D4, Div II Oxygen Concentration High-High.

ARP 17D12, Div II Oxygen Concentration High.

ILO 2018 NRC Exam Page: 71 of 282 Question 27

NUREG 1123 KA Catalog Rev. 2 500000 High Containment Hydrogen Concentration.

500000 EA2. Ability to determine and / or interpret the following as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS:

500000 EA2.04 3.3/3.3 Combustible limits for wetwell 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Primary Containment Monitoring System (T5000)

Cognitive Enabler Given various system operating parameters, relate system/equipment operation to fundamental concepts to determine proper operation/response as described in the BWR Fundamentals Catalog.

ILO 2018 NRC Exam Page: 72 of 282 Question 27

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 73 of 282 Question 28

28 K/A Importance: 4.2/4.6 Points: 1.00 R28 Difficulty: 4.00 Level of Knowledge: High Source: MODIFIED 73591 Two minutes after a small steam leak develops inside the drywell from the B Main Steam line, the following conditions exist:

• Drywell pressure is 3.0 psig.

• Reactor pressure is 750 psig and slowly lowering.

• RPV level is 170 inches and slowly rising.

Which of the following indications are correct for the current plant conditions?

A.

B.

C.

D.

A. A.

B. B.

C. C.

D. D.

Answer: D ILO 2018 NRC Exam Page: 74 of 282 Question 28

Answer Explanation:

The conditions presented in the stem indicate that RHR LPCI Loop Select Logic will have determined the loop selected for injection since a High Drywll Pressure condition (>1.68 psig) exists. Since the source of the high drywell pressure is from a Main Steam Line, Recirc Riser DP, for both loops, will be unaffected.

This will result in LPCI Loop Select Logic selecting the B Loop (Division 2 RHR) for injection because Loop A Recirc Riser DP will not be greater than Loop B Recirc Riser DP by 0.627 psig.

Since B loop is selected for injection, the A loop injection valves (E11-F015A & F017A) will get a close signal that seals in for 10 minutes, so this light will be lit on the Div 1 side. Since RPV pressure is not yet less than 461 psig (at 750 psig as stated in the stem of the question), LPCI Loop Select Logic will not have sent an open signal to the Loop B injection valves (E11-F015B & F017B) so the lights for this signal will NOT be lit.

Answer D shows the light arrangement for the conditions described above.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor is plausible if the candidate recognized correctly that Loop B was selected for injection, thereby resulting in the E11-F017A & F015A receving a closed signal. The candidate could incorrectly assume that the selected loop (Loop B) injection valves (F017B & F015B) receive an open signal (as indicated by the open light lit) at the time of selection, instead of when RPV pressure drops <461 psig.

This assumption makes this distractor incorrect.

B. This distractor is plausible if the candidate incorrectly determined that LPCI Loop Select would select the A Loop for injection based on the conditions given in the stem of the question. This is possible if the candidate incorrectly recalled the logic for determining the intact loop (comparison of riser DP) or possibly if the candidate misread the stem and thought the leak was from the B Recirc Loop instead of the B Main Steam Line. This assumption is incorrect as described above.

C. This distractor is plausible if the candidate incorrectly determined that LPCI Loop Select would select the A Loop for injection based on the conditions given in the stem of the question. This is possible if the candidate incorrectly recalled the logic for determining the intact loop (comparison of riser DP) or possibly if the candidate misread the stem and thought the leak was from the B Recirc Loop instead of the B Main Steam Line. This assumption is incorrect as described above. The candidate could incorrectly assume that the selected loop (Loop A) injection valves (F017A & F015A) receive an open signal (as indicated by the open light lit) at the time of selection, instead of when RPV pressure drops

<461 psig. This assumption is incorrect as described above.

Reference Information:

23.205, Residual Heat Removal System, Section 9.0 Emergecy Operations 23.601, Instrument Trip Sheets, Enclosure B, Logic Sheet for LPCI Loop Select Logic.

Plant Procedures 23.205 NUREG 1123 KA Catalog Rev. 2 203000 A3. Ability to monitor automatic operations of the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) including:

203000 A3.07 4.2*/4.6* Loop selection: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High MODIFIED RO ILO 2018 NRC Exam Page: 75 of 282 Question 28

Associated objective(s):

Cycle 15-4 Objectives Cognitive Terminal Given a copy of 23.205, discuss licensed operator actions to Force Low Pressure Coolant Injection (LPCI) Loop Select Logic to Division I(II)

ILO 2018 NRC Exam Page: 76 of 282 Question 28

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 77 of 282 Question 29

29 K/A Importance: 3.2/3.2 Points: 1.00 R29 Difficulty: 3.00 Level of Knowledge: High Source: NEW 73826 The plant is shut down with the following conditions:

• RHR Pump B is operating in Shutdown Cooling Mode.

Following a trip of RHRSW Pump B the CRS directs you to establish the proper RHR Service Water flow. Which of the following flow indications meets the requirements of 23.208, RHR Complex Service Water Systems?

ILO 2018 NRC Exam Page: 78 of 282 Question 29

A. B. C D A. A B. B C. C D. D Answer: C ILO 2018 NRC Exam Page: 79 of 282 Question 29

Answer Explanation:

This question tests the candidates ability to monitor and control RHR Heat Exchanger cooling flow (RHRSW flow), within allowable limits, for conditions resulting from a tripped RHRSW pump.

23.208 has a precaution that states "When only one RHR Service Water Pump is in operation with E1150-F068A (B), Div 1 (2) RHR Hx Serv Wtr Outlet FCV open, RHR Service Water flow should be maintained 5400 gpm to 6300 gpm to prevent excessive vibration of E1150-F068A (B), and to prevent pump runout".

The conditions in the stem of the question show Division 2 RHRSW in operation, with 2 pumps running, and the E1150-F068B fully open.

Upon a trip of one RHRSW pump, the candidate should recognize the need to lower flow (by throttling the E1150-F068) to protect the pump from runout and protect the valve from excessive vibration. Answer B shows RHRSW flow lowered to within the allowable band.

Distractor Explanation:

A. This distractor is plausible because the candidate could determine that RHRSW is not permitted to operate with only one RHRSW pump. Therefore, the candidate could conclude that the RHRSW flow indicator would indicate 0 gpm after the requirements of 23.208 were met. This is incorrect because one-pump RHRSW operation is permitted as long as flow is maintained within the prescribed band specified by the precaution.

B. This distractor is plausible because the indicated flow is approximately half of the flow prior to the pump trip. If the candidate assumed that flow with one pump should be about half of what it was for two pumps, then he/she could determine this is the proper ending flow.

D. The flow shown for this distractor is the indicated flow that results from tripping one RHRSW pump and not taking any operator actions, such as throttling the E1150-F068B. This distractor is plausible if the student recognized this as being the expected flow that he/she would see and if the candidate failed to recall the flow range specified in the precaution above and therefore the need to lower flow.

Reference Information:

23.208 RHR Complex Service Water Systems SOP NUREG 1123 KA Catalog Rev. 2 205000 RHR Shutdown Cooling Mode 205000 A4. Ability to manually operate and/or monitor in the control room:

205000 A4.11 3.2/3.2 Heat exchanger cooling flow 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High NEW RO ILO 2018 NRC Exam Page: 80 of 282 Question 29

Associated objective(s):

Residual Heat Removal Service Water Cognitive Enabler Identify relationships between significant Residual Heat Removal Service Water System operating parameter values.

Residual Heat Removal Service Water Cognitive Enabler Describe general Residual Heat Removal Service Water System operation, including component operating sequence, normal operating parameters, and expected system response.

Residual Heat Removal Service Water Cognitive Enabler State major precautions and limitations, and major safety considerations for the Residual Heat Removal Service Water System, and describe their bases.

ILO 2018 NRC Exam Page: 81 of 282 Question 29

30 K/A Importance: 4.6/4.6 Points: 1.00 R30 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED 73606 The HPCI system is in standby.

With the lineup shown, under which of the following conditions must HPCI be considered INOPERABLE?

A. If it remains in this configuration for more than one hour.

B. If it remains in this configuration for more than 12 consecutive hours.

C. It is currently INOPERABLE if the keep fill system is NOT in operation.

D. If it stays in this configuration for more than 12 consecutive hours, without keep fill in operation.

Answer: D ILO 2018 NRC Exam Page: 82 of 282 Question 30

Answer Explanation:

This question is RO level of knowledge because it requires the candidate to recall a procedure step (P&L) that relates to the impact of HPCI being aligned to the non-preferred suction source, or the Torus. This is required by objective LP-OP-315-0139-C002, State major precautions and limitations, and major safety considerations for the High Pressure Coolant Injection System, and describe their bases.

23.202, Step 3.19 states "If aligned to the Torus (in standby) for more than twelve consecutive hours without HPCI Keep Fill System in operation, HPCI should be considered inoperable, due to potential drain down of system piping. Reference CARD 98-11671 and TMPE 97-0345. The twelve hour clock can be reset by valving in the CST and venting the discharge line high point vent. If aligned for more than one hour on torus suction a high point vent when suction returned to CST or keep fill restored should be performed.

The conditions (picture) in the stem of the question show the HPCI system in a standby lineup with suction aligned to the Torus. Under this condition, the HPCI system is OPERABLE, however 23.202, Step 3.19 applies. This question requires the candidate to evaluate the system lineup and then interpret and execute 23.202, Step 3.19 to determine system OPERABILITY.

Answer D is correct because, given the current HPCI system configuration, if keep fill became unavailable, the system would be INOPERABLE in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor is plausible if the candidate did not correctly interpret and apply another step in 23.202, that relates to the configuration shown in the picture. 23.202 Step 3.21 states: When HPCI is aligned to the Torus without HPCI Keep Fill System in operation for greater than one hour:

• There is a potential for a void to form causing a pressure transient with HPCI startup.

• When HPCI Suction is realigned to CST, the discharge line high point vent should be vented.

• HPCI will not be considered inoperable in this condition (CARD 01-20890).

The candidate could incorrectly recall this step and consider HPCI to be INOPERABLE after one hour with its suction aligned to the Torus as shown.

B. This distractor is plausible if the candidate incorrectly interprets and applies 23.202, Step 3.19 and determines that HPCI is INOPERABLE if it remains in the configuration shown, for greater than 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, regardless of the status of the keepfill system.

C. This distractor is plausible if the candidate incorrectly interprets and applies the steps regarding the status of the HPCI system, with its suction aligned to the Torus without keep fill in operation, and determines that HPCI is INOPERABLE any time the HPCI suction is aligned to the Torus (as shown) without keep fill in operation. This is incorrect because there are conditions that allow HPCI to remain OPERABLE, with its suction aligned to the Torus and without keep fill in operation, as long as the appropriate actions are taken as specified in steps 3.19 and/or 3.21.

Reference Information:

23.202 HPCI System, Step 3.19 NUREG 1123 KA Catalog Rev. 2 206000 HPCI System.

G2.1.20 Ability to interpret and execute procedure steps 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

High MODIFIED NRC Early Review RO ILO 2018 NRC Exam Page: 83 of 282 Question 30

Associated objective(s):

High Pressure Coolant Injection Cognitive Enabler State major precautions and limitations, and major safety considerations for the HPCI System, and describe their bases.

High Pressure Coolant Injection Cognitive Enabler State major precautions and limitations, and major safety considerations for the HPCI System, and describe their bases.

ILO 2018 NRC Exam Page: 84 of 282 Question 30

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 85 of 282 Question 31

31 K/A Importance: 3.7/3.8 Points: 1.00 R31 Difficulty: 3.00 Level of Knowledge: High Source: NEW 73607 The plant is operating at 100% power with Core Spray Pumps A and C running for surveillance testing Division 1 Core Spray System Pump and Valve Operability test.

Which of the responses below best completes the following statement to describe how the Core Spray pumps would respond to a simultaneous Loss of Offsite Power (LOP) and Loss of Coolant Accident (LOCA) under these conditions?

Core Spray Pumps A and C would ____(1)____ their respective EDG output breakers closed.

Core Spray Pumps B and D would ____(2)____ their respective EDG output breakers closed.

A. (1) start five seconds after (2) start five seconds after B. (1) start five seconds after (2) immediately restart when C. (1) immediately restart when (2) start five seconds after D. (1) immediately restart when (2) immediately restart when Answer: C ILO 2018 NRC Exam Page: 86 of 282 Question 31

Answer Explanation:

Note: This explanation refers to the operation of Core Spray Pump A (see I-2211-01). All Core Spray Pumps operate in the same manner.

If power is available to bus 64B, the K3A relay will energize, closing contacts K3A in the CSS Pump A control string. K16A Time Delay Pickup (TDPU) relay will energize after 5 seconds, closing contacts K16A. This will energize relay K12A, which will start the CSS Pump 5 seconds AFTER the LOCA signal is received. Note: This 5 second delay in CSS pump starting comes from Core Spray logic and NOT the EDG Load Sequencer.

On a LOP/LOCA with Core Spray Pump A in standby, as soon as the EDG Automatic Load Sequencer (EDGALSS) senses the EDG output breaker closing, it energizes relay XK-34, which permits the Core Spray Pump breaker to close. The Core Spray pump breaker normally still has to wait on a signal from Core Spray Logic, via relay K12A (from I-2215-02), to close when in Auto.

However, when Core Spray Pump A is in RUN, as stated in the stem of the question, this 5-second time delay is effectively bypassed. In this case, when the LOP/LOCA occurs with Core Spray Pump A in RUN, as soon as the EDG Automatic Load Sequencer (EDGALSS) senses the EDG output breaker closing, it energizes relay XK-34, which permits the Core Spray Pump breaker to close. With the CMC in RUN, the Core Spray pump breaker will immediately close without needing to wait on a signal from Core Spray Logic (effectively bypassing the 5-second time delay).

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor is plausible if the candidate did not recall the difference in the response of the Core Spray system when in Automatic compared to when the pump CMC is in RUN. This is incorrect, however, as described above.

B. This distractor is plausible if the candidate recalled that the Core Spray pumps respond the same as the RHR pumps and they normally start immediately after the EDG restores power to the bus following a LOP/LOCA when in Auto. The candidate could also could recall that Core Spray logic changes when the pump is in RUN and incorrectly assume that the way it changes is to cause a 5-second time delay on start when in RUN. This is incorrect as described above.

D. This distractor is plausible if the candidate recalled that the Core Spray pumps always started immediately after the EDG restored bus power following a LOP/LOCA. This is plausible because this is how the RHR pumps respond. Since the RHR pumps restart immediately after the EDG output breakers close, following a LOP/LOCA, the candidate could incorrectly assume that the Core Spray pumps do as well since they are both low pressure injection systems. This is incorrect, however, as described above.

Reference Information:

23.203, Core Spray System I-2215-02 Div 1 and Div 2 CSS Logic Schematic I-2211-01 Core Spray Pump A Schematic ST-OP-315-0040, Core Spray System Student Text NUREG 1123 KA Catalog Rev. 2 209001 K1. Knowledge of the physical connections and/or cause effect relationships between LOW PRESSURE CORE SPRAY SYSTEM and the following:

209001 K1.10 3.7/3.8 Emergency generator 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High NEW RO ILO 2018 NRC Exam Page: 87 of 282 Question 31

Associated objective(s):

Core Spray System Cognitive Enabler Describe general Core Spray System operation, including component operating sequence, normal operating parameters, and expected system response.

Core Spray System Cognitive Enabler Describe general Core Spray System operation, including component operating sequence, normal operating parameters, and expected system response.

ILO 2018 NRC Exam Page: 88 of 282 Question 31

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 89 of 282 Question 32

32 K/A Importance: 3.1/3.2 Points: 1.00 R32V2 Difficulty: 4.00 Level of Knowledge: High Source: BANK 75086 The plant is operating at 100% power. The control power fuse for Motor Control Center (MCC) 72E-5B Pos 2B, which provides power to ONE of the SLC Pump control circuits, has blown.

Which of the responses below accurately completes the following statement regarding the status of the SLC system?

Currently, only the __(1)__ squib circuit continuity light is lit and __(2)__ squib valve(s) would open upon system initiation.

A. A; both B. B; both C. A; only the A D. B; only the B Answer: C ILO 2018 NRC Exam Page: 90 of 282 Question 32

Answer Explanation:

The A and B SLC Pumps are powered from MCCs 72B-4C, Pos 2A-R and 72E-5B, Pos 2B respectively.

Note: This question tests a common misconception among licensed operator candidates in that, regardless of what SLC pump is started, both explosive (squib) valves will always open. What actually happens is that, when either SLC pump is started, contacts in the control circuits for both squibs close that would enable both squibs to fire. However, control power is still needed to fire the squib and, since control power comes from the A and B SLC pump control circuits, both squibs will only fire if both pumps have power via their respective MCCs with control power available.

Since the stem of the question states that control power to MCC 72E-5B, Pos 2B has been lost, the candidate should recognize that power to the B SLC Pump, and B explosive (squib) control circuits, has been lost. This will cause a loss of continuity through the B squibs and loss of continuity indication (amber light). Upon system initiation, the B SLC pump will be unavailable. Upon starting the A SLC pump, even though a contact will close in the control circuit for the B squib valve (C4104F004B), without control power available, this squib will not fire and the valve will not open. Only the A squib will fire for the A squib valve (C4104F004A) and only the A valve will open.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could recognize that 72E-5B is the power supply to the B squib ciruit and, therefore, recognize that only the A squib continuity light will be lit. The candidate could then conclude that, regardless of this failure, both squib circuits will fire and both squib valves will open, which is plausible because of the common misconception that both squibs will always fire, regardless of which SLC pump is started. This is incorrect because each squib valve still requires control power from the MCC that powers the valve's respective SLC pump. The candidate could apply this misconception and assume that both valves will still open upon system initiation, which is incorrect.

B. The candidate could incorrectly determine that 72E-5B is the power supply to the A squib ciruit and, therefore, determine that only the B squib continuity light will be lit, which is incorrect because 72E-5B is the power supply to the B squib circuit. Furthermore, this distractor is plausible because of the common misconception that both squibs will always fire, regardless of which SLC pump is started.

This is incorrect because each squib valve still requires control power from the MCC that powers the valve's respective SLC pump. The candidate could apply this misconception and assume that both valves will still open upon system initiation, which is incorrect.

D. This distractor is plausible because the operator could incorrectly determine that the MCC given in the stem of the question (72E-5B, Pos 2B) powers the A SLC Pump and A squib control circuit. This is not true since the A SLC Pump and squib valve control circuits are powered from 72B-4C, Pos 2A-R.

Reference Information:

I-2131-01, SLC Pumps Schematic Diagram Student Text ST-OP-315-0014, SLC System NUREG 1123 KA Catalog Rev. 2 211000 SLC System 211000 K2. Knowledge of electrical power supplies to the following:

211000 K2.02 3.1*/3.2* Explosive valves 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK High RO ILO 2018 NRC Exam Page: 91 of 282 Question 32

Associated objective(s):

Standby Liquid Control Cognitive Terminal Given various controls and indications for Standby Liquid Control, analyze that indication to determine proper plant response in accordance with approved plant procedures.

ILO 2018 NRC Exam Page: 92 of 282 Question 32

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 93 of 282 Question 33

33 K/A Importance: 3.7/3.9 Points: 1.00 R33 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED 73567 What is the effect that a loss of RPS Bus B will have on the RHR system with Division 1 RHR running in Shutdown Cooling on RHR Pump A?

A. RHR will continue to operate, but some Division 1 RHR instrumentation will be INOPERABLE.

B. E1150-F008, RHR SDC Otbd Suction Iso Vlv, will close resulting in a loss of Shutdown Cooling.

C. E1150-F009, RHR SDC Inbd Suction Iso Vlv, will close resulting in a loss of Shutdown Cooling.

D. E1150-F008, RHR SDC Otbd Suction Iso Vlv, AND E1150-F009, RHR SDC Inbd Suction Iso Vlv, will close resulting in a loss of Shutdown Cooling.

Answer: B ILO 2018 NRC Exam Page: 94 of 282 Question 33

Answer Explanation:

As stated in 23.316, Enclosure B, a loss of RPS Bus B will cause the E1150-F008 to close. This will cause any RHR pumps running in SDC to trip on loss of suction interlock and an ultimate loss of SDC.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This answer is plausible if the candidate incorrectly assumed that, since RHR was running in SDC on Division 1, a loss of the RPS B power supply (a Division 2 power supply) would have no impact.

Additional plausibility is lent to this distractor since the second part of the distractor is correct because the E11-F080 A & B (Div 1 & Div 2 RHR Heat Exch Outlet Sample Otbd Iso Vlvs) close, which isolates Tech Spec instruments if SDC sample pts 42a or 42b are in service.

C. This distractor is plausible if the candidate fails to recognize the relationship between RPS B and the E1150-F008 and instead incorrectly recalls that RPS B loss will cause the E1150-F009 to close. This plasibility is even greater considering the somewhat confusing numbering relationship with the RHR valves in that the inboard valve is the F009 and the outboard valve is the F008. Because of this, licensed operator candidates often confuse the numbering of these valves. If the candidate got distracted by the confusing numbering scheme, as often happens, he/she could incorrectly conclude that RPS B impacts the E1150-F009.

D. This distractor is plausible if the candidate incorrectly recalled that a loss of either RPS Bus caused both RHR SDC containment isolation valves to go closed, which is the case for some PCIS group isolation valves, such as the T4901-F602 and T4901-F468, Division 2 DW Pneumatics Inbd and Otbd Iso Valves, which both go closed on a loss of RPS B. This is incorrect, however, since only the E1150-F008 will close on a loss of RPS B.

Reference Information:

23.316, RPS 120Vac and RPS MG Sets, Enclosure B (RPS Bus B - Affected Equipment List)

Student Text ST-OP-315-0027, RPS System NUREG 1123 KA Catalog Rev. 2 212000 RPS 212000 K3. Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION SYSTEM will have on the following:

212000 K3.02 3.7/3.9 Primary containment isolation system/nuclear steam supply shut-off: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High MODIFIED RO Associated objective(s):

Reactor Protection System (C7100)

Cognitive Enabler Discuss the Reactor Protection System interrelationships with other systems.

Reactor Protection System (C7100)

Cognitive Enabler Discuss the Reactor Protection System interrelationships with other systems.

ILO 2018 NRC Exam Page: 95 of 282 Question 33

34 K/A Importance: 3.7/3.7 Points: 1.00 R34 Difficulty: 2.00 Level of Knowledge: High Source: BANK 73568 The RETRACT PERMIT light for IRM G is NOT lit. Which of the following describes the effect on IRM G?

A. It will NOT retract.

B. It can be retracted. Doing so will cause a Rod Block.

C. It can ONLY be retracted if it is on Range 1. Doing so will cause a Rod Block.

D. It can ONLY be retracted if it is on Range 1. Doing so will NOT cause a Rod Block.

Answer: B ILO 2018 NRC Exam Page: 96 of 282 Question 34

Answer Explanation:

The IRMs are not normally withdrawn until in MODE 1. Upon entering MODE 1, the Mode Switch is taken to RUN at which time all IRM RETRACT PERMIT lights will come on. The RETRACT PERMIT light is simply information to the operator that conditions are met to withdraw a detector. The RETRACT PERMIT circuitry does not permit or block actual withdrawal of a detector. Without a RETRACT PERMIT, if an unbypassed IRM detector is not fully inserted, a Rod Block will ensue. Therefore, under the conditions provided in the stem of the question, the operator could retract the IRM detector, however, doing so will result in a Rod Block.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor is plausible because the candidate could take the name of the light (RETRACT PERMIT) literally and assume that, without the light lit, withdrawal of the detector would not be possible. However, this is not true because the RETRACT PERMIT circuitry does not permit or block actual withdrawal of a detector.

C. This distractor is plausible because of an interrelationship that exists between the IRMs and SRMs. If the associated IRMs are all on Range 1, then an SRM can go below 100 cps without causing a Rod Block. If the candidate misapplied the Range 1 interrelationship to the conditions given in the stem of this question, he/she could conclude that the IRM can only be retracted if on Range 1. This is incorrect because the IRMs will retract on all ranges. The candidate may recognize that retracting the IRM, however, without the RETRACT PERMIT light will cause a Rod Block, which is true.

D. This distractor is plausible because of an interrelationship that exists between the IRMs and SRMs. If the associated IRMs are all on Range 1, then an SRM can go below 100 cps without causing a Rod Block. If the candidate misapplied the Range 1 interrelationship to the conditions given in the stem of this question, he/she could conclude that the IRM can only be retracted if on Range 1. This is incorrect because the IRMs will retract on all ranges. The candidate may also incorrectly conclude that, because the IRM is on Range 1, a Rod Block will not occur (similar to how the SRM rod block is prevented with the associated IRMs are on Range 1).

Reference Information:

I-2145-56 SRM IRM Aux Relays I-2145-59 NMS RMCS-RPS Interlocks 23.602 SRM System 23.603 IRM System Student Text ST-OP-315-0022, SRM System Student Text ST-OP-315-0023, IRM System NUREG 1123 KA Catalog Rev. 2 215003 K4. Knowledge of INTERMEDIATE RANGE MONITOR (IRM) SYSTEM design feature(s) and/or interlocks which provide for the following:

215003 K4.01 Rod withdrawal blocks 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK High RO ILO 2018 NRC Exam Page: 97 of 282 Question 34

Associated objective(s):

Intermediate Range Monitoring (C5111)

Cognitive Enabler List the interlocks associated with Intermediate Range Monitoring System components.

Intermediate Range Monitoring (C5111)

Cognitive Enabler List the interlocks associated with Intermediate Range Monitoring System components.

ILO 2018 NRC Exam Page: 98 of 282 Question 34

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 99 of 282 Question 35

35 K/A Importance: 2.6/2.6 Points: 1.00 R35 Difficulty: 3.00 Level of Knowledge: High Source: NEW 73646 How is the Signal-to-Noise ratio of an SRM detector determined and what is the operational implication of this number?

A.

If it is 20:1, then the minimum allowed count rate is 0.7 cps.

B.

If it is 20:1, then the minimum allowed count rate is 0.7 cps.

C.

If it is 20:1, then the minimum allowed count rate is 3.0 cps.

D.

If it is 20:1, then the minimum allowed count rate is 3.0 cps.

Answer: A ILO 2018 NRC Exam Page: 100 of 282 Question 35

Answer Explanation:

With an SRM detector fully inserted, it is in a high neutron flux region of the core. This is the core location where most of the signal produced will be the result of neutron interactions. At this location, the detector output will be a combination of both the neutron signal plus any background noise.

With an SRM detector fully withdrawn, it is in a low flux region (approximately 2.5 feet below the bottom of active fuel) where very little, if any, of the detector output will be the result of neutron interactions. At this location, the detector output will be the result of background noise only.

These two signals can be used to check for conformance with the signal-to-noise ratio requirement.

The candidate must determine that, with the detector fully inserted, the resultant signal output equals signal + noise. When withdrawn the output is just noise. If the withdrawn value is subtracted from the insert value, the result is known as the 'signal'. Then, if the resulting signal is divided by the withdrawn value (noise), the end result is known as the signal to noise ratio.

From Technical Specifications Section 3.3.1.2, a detector is considered to be OPERABLE as long as it reads at least 3.0 cps OR if it reads at least 0.7 cps as long as the detector's signal-to-noise ratio (determined above) is at least 20:1.

This information, in addition to being a TS requirement, is also a requirement of the Reactor Startup GOP (General Operating Procedure) that the RO uses when performing a Reactor Startup, specifically GOP Step 22.000.02, Step 5.2.7.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. This distractor is plausible if the candidate did not fully understand SRM detector operation and assumed that the detector 'signal' is determined with the SRM withdrawn from the core. This is incorrect since the SRMs are miniature fission chambers that develop their 'signal' from interactions with thermal neutrons.

C. This distractor is plausible if the candidate determined the correct method for calculating the signal-to-noise ratio but mis-applied the Tech Spec Surveillance requirement and assumed that a ratio greater than 20:1 required a higher count rate of at least 3.0 cps, which is incorrect.

D. This distractor is plausible if the candidate did not fully understand SRM detector operation and assumed that the detector 'signal' is determined with the SRM withdrawn from the core. This is incorrect since the SRMs are miniature fission chambers that develop their 'signal' from interactions with thermal neutrons. The candidate could also incorrectly apply the Tech Spec Surveillance requirement and assume that a ratio greater than 20:1 required a higher count rate of at least 3.0 cps, which is incorrect.

Reference Information:

ST-OP-315-0022, SRM System Student Text.

24.000.01, Situational Surveillances/LCO Action Tracking, Attachment 37.

Tech Spec Bases for SR 3.3.1.2.5 and 3.3.1.2.6.

NUREG 1123 KA Catalog Rev. 2 215004 K5. Knowledge of the operational implications of the following concepts as they apply to SOURCE RANGE MONITOR (SRM) SYSTEM:

215004 K5.01 2.6/2.6 Detector operation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Question Use (ILO 2018)

High NEW RO ILO 2018 NRC Exam Page: 101 of 282 Question 35

Associated objective(s):

Source Range Monitoring (C5110)

Cognitive Enabler Discuss design considerations, capabilities, and limitations related to Source Range Monitoring System component operation.

Reactor Startup Certification Performance Terminal Perform a reactor startup in accordance with 22.000.02, Plant Startup to 25% Power, from reactor shutdown to establishing a <90°F/Hr heatup rate Source Range Monitoring (C5110)

Cognitive Enabler Describe general Source Range Monitoring System operation, including component operating sequence, normal operating parameters, and expected system response.

Source Range Monitoring (C5110)

Cognitive Enabler Describe general Source Range Monitoring System operation, including component operating sequence, normal operating parameters, and expected system response.

ILO 2018 NRC Exam Page: 102 of 282 Question 35

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 103 of 282 Question 36

36 K/A Importance: 3.1/3.2 Points: 1.00 R36 Difficulty: 3.00 Level of Knowledge: High Source: NEW 74966 The plant was operating at 100% power when APRM #1 failed upscale.

BEFORE any operator actions are taken, the Recirculation Flow Unit for APRM #3 subsequently fails downscale.

Which of the following responses describes the impact of these malfunctions on the APRM System 2/4 Logic Modules?

A. ONLY 2/4 Logic Module #1 will have a trip from APRM #1.

B. ALL four 2/4 Logic Modules will have a trip from APRM #1 ONLY.

C. ALL four 2/4 Logic Modules will have trips from BOTH APRM #1 and APRM #3.

D. 2/4 Logic Module #1 will have a trip from APRM #1 ONLY and 2/4 Logic Module

1. 3 will have a trip from APRM #3 ONLY.

Answer: C ILO 2018 NRC Exam Page: 104 of 282 Question 36

Answer Explanation:

This question requires the candidate to determine the effect of multiple APRM failures on the overall APRM System by determining the trips that will be present on the APRM 2/4 Logic Modules. Note: The 2/4 Logic Modules are the interface between the NUMAC APRM System and the Reactor Protection System (RPS) and are also known as Voters.

The candidate should determine that the first APRM upscale failure will result in each of the 4 APRM Voters receiving one "vote" from APRM #1. The candidate should then determine that the subsequent downscale failure of the flow input to APRM #3 will result in APRM #3 calculating a Simulated Thermal Power (STP) trip (flow-biased trip) setpoint that is below actual APRM Power, which will result in a trip output from APRM #3 as well. This second trip will also get "voted" by all 4 APRM voters. Therefore, the candidate should conclude that all 4 APRM 2/4 Logic Modules will see trips from both APRMs #1 and #3.

Although not asked, the overall result of these two failures, and two "votes", is that all 4 Voters will go to a tripped condition, which will cause both RPS A and B to trip, resulting in a reactor scram.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly recall that the 4 APRMs only output to their associated Voters, which could lead the candidate to determine that the upscale trip from APRM #1 will only be seen by Voter

1. 1, which is incorrect because all 4 APRMs input into all 4 Voters. This misconception is often caused by licensed operator candidates misunderstanding the relationship between the 2/4 Logic Modules and their 'home' APRM. Each Voter has a 'home' APRM (#1 for #1, #2 for #2, and so on) that it receives bypass and self-test information from. However, this does not change the fact that each Voter receives trip input information from ALL 4 APRMs and NOT just their 'home' APRM. The candidate could also fail to determine that the downscale failure of the flow input to APRM #3 will cause it to trip on STP Upscale.

B. The candidate could correctly recall that all 4 APRMs input into all 4 Voters and determine that the Upscale Trip from APRM #1 will be seen by all 4 Voters. However, the candidate could fail to recognize the significance of the downscale failure of the flow input to APRM #3 and not conclude that APRM #3 will also be in a tripped condition due to a STP Upscale Trip, which is incorrect because the downscale failure of the flow input will result in a low STP Setpoint, compared to actual power, which will put APRM #3 in a tripped condition.

D. The candidate could incorrectly recall that the 4 APRMs only output to their associated Voters, which could lead the candidate to determine that the upscale trip from APRM #1 will only be seen by Voter

1. 1, and the Upscale Trip from APRM #3 will only be seen by Voter #3, both of which are incorrect because all 4 APRMs input into all 4 Voters. Again, this misconception is often caused by candidates misunderstanding the relationship between the Voters and their 'home' APRMs as described in A above.

Reference Information:

23.605 APRM System SOP Student Text ST-OP-315-0024, PRNM System NUREG 1123 KA Catalog Rev. 2 215005 APRM/LPRM 215005 K6. Knowledge of the effect that a loss or malfunction of the following will have on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM:

215005 K6.04 3.1/3.2 Trip units 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High NEW RO ILO 2018 NRC Exam Page: 105 of 282 Question 36

Associated objective(s):

Power Range Neutron Monitoring (C5112, C5113 & C5114)

Cognitive Terminal Given various controls and indications for PRNM operations, analyze that indication to determine proper plant response in accordance with approved plant procedures.

ILO 2018 NRC Exam Page: 106 of 282 Question 36

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 107 of 282 Question 37

37 K/A Importance: 3.7/3.7 Points: 1.00 R37 Difficulty: 3.00 Level of Knowledge: High Source: BANK 73706 The Reactor Core Isolation Cooling (RCIC) system has been started in accordance with 24.206.04, RCIC SYSTEM AUTOMATIC ACTUATION AND FLOW TEST.

The following conditions exist:

• RCIC Flow Controller, E51-K615, is in Auto, set at 650 gpm

• RCIC flow - 650 gpm

• RPV pressure - 185 psig What happens to RCIC final speed and discharge pressure if E41-F011 HPCI Test Line Iso/PCV is OPENED an additional 5%?

A. RCIC speed lower RCIC discharge pressure lower B. RCIC speed lower RCIC discharge pressure higher C. RCIC speed higher RCIC discharge pressure lower D. RCIC speed higher RCIC discharge pressure higher Answer: A ILO 2018 NRC Exam Page: 108 of 282 Question 37

Answer Explanation:

The RCIC system is tested during plant operation by recirculating condensate tank (CST) water through the HPCI/RCIC test return line. A drag type valve (E41-F011) is employed in the test return line to simulate discharge pressure conditions. E41-F011 is designed to simulate conditions where the RCIC system is supplying cooling water to the reactor vessel under high and low reactor vessel pressure. Required test differential pressure is obtained by use of a differential pressure controller which maintains the test pressure at the desired setpoint. E41-F011 provides a differential pressure to demonstrate RCIC will provide the required flow rate at simulated reactor pressures of 1135 psia and 135 psia. The simulated pressure represents reactor pressure plus pressure drop from line losses between RCIC and the reactor vessel. This valve is also used for developing the required HPCI differential test pressure.

When the valve is opened, as stated in the stem of the question, RCIC pump back pressure will lower, which will cause flow through the test line to increase and RCIC system flow to increase. With the RCIC flow controller (E51-K615) in automatic, as stated in the stem of the question, the increase in system flow will be sensed by the controller which will close down on the RCIC trip throttle valve to restore system flow back to the setpoint of 650 gpm. This will result in less steam flow to the RCIC turbine, thereby lowering turbine speed.

The end result will be system flow restored to the setpoint of 650 gpm, at a lower discharge pressure (because of less backpressure and reduced headloss by opening the drag valve) and lower turbine speed.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could determine that the test line pressure control valve directly controls RCIC pump discharge pressure (rather than acting as a backpressure control, or drag valve) and therefore assume that opening it would result in a higher discharge pressure. If this was assumed, the candidate could then determine that the higher discharge pressure would result in more system flow, so the turbine would have to slow down to bring flow back to the controlled setpoint.

C. The candidate could recognize that opening the test line PCV could lower system back pressure and, therefore, pump discharge pressure, which is correct. The candidate could assume that the turbine speed would have to increase to develop this higher pressure, which is not true. The candidate could fail to recognize the impact of the reduced backpressure on system flow and the response of the flow controller to lower turbine speed in an effort to control system flow at the setpoint.

D. The candidate could determine that the test line pressure control valve directly controls RCIC pump discharge pressure (rather than acting as a backpressure control, or drag valve) and therefore assume that opening it would result in a higher discharge pressure. This could lead the candidate to determine that the turbine speed would have to increase to develop this higher pressure, which might be true if it was a discharge pressure control valve, but it is not true as described above.

Reference Information:

24.206.04, RCIC System Automatic Actuation and Flow Test Student Text ST-OP-315-0043, RCIC System GE BWR Fundamentals, Components, Chapter 2, Pumps GE BWR Fundamentals, Thermodynamics Chapter 6, Fluid Statics and Dynamics.

E51-00 HPCI System Design Basis Document (DBD). Note: The E41-F011 is a test valve shared between the RCIC and HPCI systems. The explanation in the HPCI DBD is better than in the RCIC DBD. Also, the RCIC DBD is coded as "confidential" due to proprietary information contained therin.

ILO 2018 NRC Exam Page: 109 of 282 Question 37

NUREG 1123 KA Catalog Rev. 2 217000 RCIC System 217000 A1. Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) controls including:

217000 A1.05 3.7/3.7 RCIC turbine speed 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Question Use (ILO 2018)

BANK High RO Associated objective(s):

Reactor Core Isolation Cooling Cognitive Enabler Given various system operating parameters, relate system/equipment operation to fundamental concepts to determine proper operation/response as described in the BWR Fundamentals Catalog.

ILO 2018 NRC Exam Page: 110 of 282 Question 37

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 111 of 282 Question 38

38 K/A Importance: 4.2/4.3 Points: 1.00 R38 Difficulty: 4.00 Level of Knowledge: High Source: BANK 73626 Following a Loss of Offsite Power, the following conditions occur at the listed times:

• 12:00 Reactor Scram occurred, all Control Rods are inserted.

• 12:01 ONLY EDG 14 has started and loaded.

• 12:05 Drywell Pressure is 0.9 psig and stable.

• 12:10 RPV Water Level is 64 inches, lowering 4 inches per minute.

Given these conditions: (1) Which ONE of the following describes the response of the Automatic Depressurization (ADS) System and (2) the operator actions that should be taken to prevent ADS initiation when directed to do so by the CRS when in the EOPs?

A. (1) ADS will OPEN Safety Relief Valves at 12:20.

(2) Place the ADS Inhibit Sw Logic A (B) key switches in INHIBIT.

B. (1) ADS will OPEN Safety Relief Valves at 12:27.

(2) Place the ADS Inhibit Sw Logic A (B) key switches in INHIBIT.

C. (1) ADS will OPEN Safety Relief Valves at 12:20.

(2) Simultaneously depress the ADS Division I (II) Timer Logic RESET pushbuttons.

D. (1) ADS will OPEN Safety Relief Valves at 12:27.

(2) Simultaneously depress the ADS Division I (II) Timer Logic RESET pushbuttons.

Answer: B ILO 2018 NRC Exam Page: 112 of 282 Question 38

Answer Explanation:

Correct Answer: B With NO High Drywell Pressure signal present, L1 (31.8 inches) will cause ADS Timer to initiate in 7 minutes. The ADS Timer lasts 105 seconds or 2 minutes. A total of 9 minutes later, SRVs will OPEN. L1 will be reached in 8 minutes. 8 minutes + 9 minutes = 17 minutes 12:10 + 17 minutes

= 12:27.

23.201, Section 5.4 gives specific direction for preventing ADS initiation by placing the ADS Inhibit Sw Logic A and B key switches in INHIBIT. These are the actions that the operator would take if directed by the CRS to inhibit ADS per the EOPs.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor is plausible because the time would be true if Drywell Presssure was above 1.68 psig.

When above 1.68 psig, the 7-minute timer is bypassed so the 105 second countdown occurs when L1 is reached in 8 minutes. 8 minutes + 2 minutes = 10 minutes. 12:10 + 10 minutes = 12:20. The second part of this distractor is correct.

C. This distractor is plausible because the time would be true if Drywell Presssure was above 1.68 psig.

When above 1.68 psig, the 7-minute timer is bypassed so the 105 second countdown occurs when L1 is reached in 8 minutes. 8 minutes + 2 minutes = 10 minutes. 12:10 + 10 minutes = 12:20. The second part is plausible because these actions would cause the ADS 105 second timer to reset and start couting down again. However, it is incorrect because these actions would not prevent ADS initiation nor are they the actions that an operator is expected to take when directed by the CRS, per the EOPs, to prevent ADS initiation. These push buttons are only used to RESET ADS logic, following initiation, in the ADS System Shutdown Section (Section 6.3) of 23.201.

D. The first part of this distractor (timing of ADS initiation) is correct. The second part is plausible because these actions would cause the ADS 105 second timer to reset and start couting down again.

However, it is incorrect because these actions would not prevent ADS initiation nor are they the actions that an operator is expected to take when directed by the CRS, per the EOPs, to prevent ADS initiation. These push buttons are only used to RESET ADS logic, following initiation, in the ADS System Shutdown Section (Section 6.3) of 23.201.

Reference Information:

23.201, SRV and ADS System, Section 5.4 Automatic Depressurization System Operation.

NUREG 1123 KA Catalog Rev. 2 218000 ADS 218000 A2. Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

218000 A2.06 ADS initiation signals present 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Question Use (ILO 2018)

BANK High RO ILO 2018 NRC Exam Page: 113 of 282 Question 38

Associated objective(s):

Automatic Depressurization System (B2104)

Cognitive Enabler Describe general ADS operation, including component operating sequence, normal operating parameters, and expected system response.

Automatic Depressurization System (B2104)

Cognitive Enabler Describe general ADS operation, including component operating sequence, normal operating parameters, and expected system response.

ILO 2018 NRC Exam Page: 114 of 282 Question 38

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 115 of 282 Question 39

39 K/A Importance: 3.4/3.4 Points: 1.00 R39 Difficulty: 3.00 Level of Knowledge: High Source: BANK 73726 The plant is at 100% power when a change in plant conditions results in the following indications:

What will the Main Steam Isolation Valve (MSIV) indicating lights show 10 seconds later?

A. All MSIV RED OPEN indicating lights will be lit.

B. All MSIV GREEN CLOSED indicating lights will be lit.

C. The B2103-F022A-D, Inboard MSIV, GREEN CLOSED indicating lights will be lit.

The B2103-F028A-D, Outboard MSIV, RED OPEN indicating lights will be lit.

D. The B2103-F022A-D, Inboard MSIV, RED OPEN indicating lights will be lit.

The B2103-F028A-D, Outboard MSIV, GREEN CLOSED indicating lights will be lit.

ILO 2018 NRC Exam Page: 116 of 282 Question 39

Answer: A ILO 2018 NRC Exam Page: 117 of 282 Question 39

Answer Explanation:

This question requires the candidate to interpret the information provided by some indicating lights associated with the Primary Containment Isolation System (PCIS) and then determine the impact of that prediction on MSIV indicating light status.

Per ARP 2D36, with channels B or D tripped, no isolation actions will occur. IF a B/D channel were to trip AND an A/C channel were to trip, then a full MSIV isolation would occur. This is known as 1 out of 2 twice logic or [Full Isolation = (A OR C) AND (B OR D)].

Therefore, A is the correct answer since only half of the trip logic was satisfied with the B and D instruments being in trip as shown in the stem of the question. Therefore, the candidate must determine that the MSIV red open indicating lights will be lit.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could incorrectly recall MSIV isolation logic as 2 out of 2 taken once or [Full isolation =

(A AND C) OR (B AND D)]. With this assumption, and with the B and D channels tripped, the candidate could determine that a full isolation has occurred.

C. The candidate could incorrectly recall MSIV isolation logic as 2 out of 2 taken once or [Otbd isolation =

(A AND C)] OR [Inbd isolation = (B AND D)]. With this assumption and the B and D channels tripped, the candidate could determine that only the Inbd MSIVs would indicate closed (green).

D. The candidate could incorrectly recall MSIV isolation logic as 2 out of 2 taken once or [Inbd isolation =

(A AND C)] OR [Otbd isolation = (B AND D)]. With this assumption and the B and D channels tripped, the candidate could determine that only the Otbd MSIVs would indicate closed (or green).

Reference Information:

ARP 2D36 NSSS Isolation Ch B/D Trip ST-OP-315-0048, PCIS System Student Text ST-OP-315-0005, Nuclear Boiler System Student Text NUREG 1123 KA Catalog Rev. 2 223002 PCIS/NSSS 223002 A3. Ability to monitor automatic operations of the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF including:

223002 A3.01 3.4/3.4 System indicating lights and alarms 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK High RO Associated objective(s):

Primary Containment Isolation System Cognitive Enabler Discuss effective monitoring of Primary Containment Isolation System using local, remote, computer displays and alarms.

ILO 2018 NRC Exam Page: 118 of 282 Question 39

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 119 of 282 Question 40

40 K/A Importance: 3.2/3.2 Points: 1.00 R40 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 73746 Following an uncomplicated reactor scram from 100% power, what is the final speed of the Reactor Recirculation Pumps and why?

A. 28%, due to DCS logic sensing both RRMG Set Field Breakers open.

B. 30%, due to the Feedwater Control System sensing less than 20% Total Feedwater Flow.

C. 37% due to the Feedwater Control System sensing RPV Level less than Level 4 with low feed pump flow .

D. 40% due to the DCS logic sensing closure of both heater drain check valves, N2200-F026A and B, for greater than 3 seconds.

Answer: B ILO 2018 NRC Exam Page: 120 of 282 Question 40

Answer Explanation:

On a reactor scram, Post Scram Feedwater Logic reduces the speed of the operating feedwater pumps in order to prevent over-feeding the reactor vessel. This reduction in feedwater flow results in total feedwater flow dropping less than 20%. This total feedwater flow signal is developed by Feedwater DCS. The Reactor Recirculation System (RRS) receives this Total Feedwater flow signal from the Feedwater Control System (FWCS) for the purpose of generating a RR Limiter #1 signal. RR Limiter 1 is actuated when total Feedwater flow is <20% (approximately 3.15 Mlbm/hr). Recirc DCS then limits RR flow to 30% speed, which is where the RR pumps end up following a scram.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. 28% speed is the minimum speed of the RR DCS controllers. When the RRMG set field breakers are open, RR DCS automatically adjusts the RRMG set speed controller setpoint to 28% in preparation for the next RRMG set startup. It is a common misconception among licensed operator candidates that the RR pumps run back to minimum (28%) speed, vice the actual 30% speed, on limiter #1 because of this design feature.

C. 37% is the nominal speed for RR Limiter #2. Limiter 2 is related to Feedwater DCS because it is generated from the receipt of an RPV Level 4 signal (low RPV water level alarm setpoint) from the Feedwater Control System in conjunction with a RFP trip OR low RFP suction flow signal. The candidate could confuse the low RFP suction flow signal (which comes in following every scram due to the reduction in Feedwater Flow) with the 20% feedwater flow signal discussed above, and conclude that the RR pumps must end up at 37% speed following a reactor scram.

D. 40% speed is the nominal speed for RR Limiter #3. Limiter 3 occurs when both Heater Drain Check Valves, N2200-F026A and B, are closed for greater than 3 seconds, which signifies a loss of heater drains. On a reactor scram, a loss of heater drain flow occurs every time, which does bring in Limiter

1. 3. The candidate could recognize this and assume that the RR pumps stop at Limiter #3 on a reactor scram. However, the RR pumps do not stop at Limiter 3 and instead lower down to the Limiter 1 setpoint of 30%.

Reference Information:

23.107, Reactor Feedwater and Condensate Systems 23.138.01, Reactor Recirculation System NUREG 1123 KA Catalog Rev. 2 259002 Reactor Water Level Control System 259002 K1 Knowledge of the physical connections and/or cause effect relationships between REACTOR WATER LEVEL CONTROL SYSTEM and the following:

259002 K1.08 3.2/3.2 Recirculation system: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Feedwater Control (C3200)

Cognitive Enabler Discuss the Feedwater Control System interrelationships with other systems.

ILO 2018 NRC Exam Page: 121 of 282 Question 40

41 K/A Importance: 3.7/4.1 Points: 1.00 R41 Difficulty: 3.00 Level of Knowledge: High Source: NEW 73928 The plant is in MODE 5. 24.404.02 Section 5.3, Division 1 SGTS and Secondary Containment Isolation Dampers 18 Month Operability Test is in progress with the following plant conditions:

• RBHVAC is in service.

• Both divisions of SGTS are in standby.

• Division 1 CCHVAC is running in the Normal mode.

• Refueling operations are in progress on RB-5.

• Underwater touchup painting is in progress inside the Torus.

What will be the impact of the operator arming and depressing the Division 1 Manual Isolation TRIP pushbutton and what action is required because of this impact?

A. Both divisions of SGTS will start, direct the Torus painting crew to stop.

B. RBHVAC will trip, notify refuel floor personnel that RBHVAC will be lost.

C. Both divisions of SGTS will start, place the Division 2 SGTS Exhaust Fan CMC in OFF-RESET.

D. CCHVAC will shift to the Recirculation Mode, start Division 2 CCHVAC and shut down Division 1.

Answer: B ILO 2018 NRC Exam Page: 122 of 282 Question 41

Answer Explanation:

The surveillance procedure has a precaution (2.3) and several notes that inform the operator of the impact of this surveillance on the RBHVAC system. Performance of this surveillance will cause RBHVAC to isolate, so the surveillance requires the operator to notify Refuel Floor personnel if the surveillance is performed during refuel floor activities, as specified in the stem of the question.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Normally, any Secondary Containment Isolation signal will cause both divisions of SGTS to automatically start. However, the manual isolation pushbuttons are divisional and allow for divisional testing of the logic, such as with the surveillance being run in the stem of this question. In this case, depressing the Division 1 Manual Isolation TRIP pushbutton will only cause a start of Division 1 SGTS.

If the candidate thought that both divisions of SGTS would start, then directing the crew to stop Torus painting could be appropriate because the surveillance procedure has a precaution (2.1) and several notes that prohibit running the surveillance with painting activities in progress in the Reactor Building.

However, Precaution 3.6 of the SOP, 23.404, has a clarification that states "torus touchup painting underwater doesn't release vapors to RB atmosphere and would be excluded from this requirement."

so the test could continue with underwater Torus touchup painting in progress.

C. Normally, any Secondary Containment Isolation signal will cause both divisions of SGTS to automatically start. However, the manual isolation pushbuttons are divisional and allow for divisional testing of the logic, such as with the surveillance being run in the stem of this question. In this case, depressing the Division 1 Manual Isolation TRIP pushbutton will only cause a start of Division 1 SGTS.

If the candidate thought that both divisions of SGTS would start, then placing the Division 2 Exhaust Fan in OFF-RESET would prevent this from occurring.

D. The surveillance procedure has a precaution (2.3) and several notes that inform the operator that the performance of Section 5.3 will cause CCHVAC to shift to the Recirculation Mode. However, CCHVAC logic causes BOTH divisions to shift to Recirc for any Secondary Containment Isolation, regardless of which division is running, due to the Control Room Emergency Filtration (CREF) function of CCHVAC. If the candidate thought that CCHVAC was divisional, then starting Div 2 and shutting down Div 1 CCHVAC would be a correct course of action.

Reference Information:

24.404.02, Division 1 SGTS Filter and Secondary Containment Isolation Damper Operability Test.

23.404, Standby Gas Treatment System SOP NUREG 1123 KA Catalog Rev. 2 261000 Standby Gas Treatment System G2.2.12 Knowledge of surveillance procedures 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Standby Gas Treatment System Cognitive Enabler State major precautions and limitations, and major safety considerations for the Standby Gas Treatment System, and describe their bases.

ILO 2018 NRC Exam Page: 123 of 282 Question 41

42 K/A Importance: 2.9/3.2 Points: 1.00 R42 Difficulty: 2.50 Level of Knowledge: High Source: BANK 73946 The reactor has scrammed due to a Loss of ALL Offsite Power (LOP).

ONLY EDGs 13 & 14 have started and loaded.

What is the status of DC electrical distribution?

A. Div 1 Chargers are supplying Div 1 DC loads.

Div 2 Chargers are supplying Div 2 DC loads.

B. Div 1 Chargers are supplying Div 1 DC loads.

Div 2 Batteries are supplying Div 2 DC loads.

C. Div 1 Batteries are supplying Div 1 DC loads.

Div 2 Chargers are supplying Div 2 DC loads.

D. Div 1 Batteries are supplying Div 1 DC loads.

Div 2 Batteries are supplying Div 2 DC loads.

Answer: D ILO 2018 NRC Exam Page: 124 of 282 Question 42

Answer Explanation:

When power is lost to the ESF 4160V and 480V busses, a load shed will occur. When the EDGs restore power to the busses, EDG load sequencing will restore power to certain safety related loads. However, the ESF battery chargers do NOT get re-sequenced back on. Although the busses supplied by EDGs 13 and 14 have power available, the battery chargers remain de-energized unless manual operator action is taken. Therefore, until the Division 2 battery chargers are restored (per 20.300.Offsite, Steps AU through AY), all station DC loads will be supplied by the station batteries.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This is the normal configuration for DC electrical distribution. If the candidate failed to recognize the impact of the AC power loss, he/she could incorrectly determine that the DC electrical distribution configuration did not change for either division.

B. This is the normal configuration for Div 1 DC electrical distribution. If the candidate failed to recognize the impact of the loss of the Div 1 EDGs or if the candidate failed to recognize that all offsite power was lost and, because only Div 2 EDGs started, assumed that Div 1 offsite power was still available, he/she could incorrectly determine that the DC electrical distribution configuration did not change. The second part of this distractor is correct.

C. The first part of this distractor is correct. For the second part, the candidate could assume that the Division 2 battery chargers are restored automatically, via load sequencing, when the Div 2 EDGs (13

& 14) are loaded. This is incorrect because the battery chargers do not automatically re-sequence as described above.

Reference Information:

20.300.Offsite, Steps AU through AY for actions taken to restore the Division 1 battery chargers.

20.300.Offsite Bases, Description of actions taken for AU.1 through AY.1.

NUREG 1123 KA Catalog Rev. 2 262001 AC Electrical Distribution 262001 K3. Knowledge of the effect that a loss or malfunction of the A.C. ELECTRICAL DISTRIBUTION will have on following:

262001 K3.03 2.9/3.2 D.C. electrical distribution 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK High RO Associated objective(s):

DC Electrical Distribution (R3200 & S3102)

Cognitive Enabler Discuss the DC Electrical Distribution System interrelationships with other systems.

ILO 2018 NRC Exam Page: 125 of 282 Question 42

43 K/A Importance: 3.1/3.4 Points: 1.00 R43 Difficulty: 3.00 Level of Knowledge: High Source: BANK 73968 UPS B local indication shows that the unit's rectifier has failed and has no output.

From where is the UPS B distribution cabinet being supplied power?

A. The UPS battery, via the UPS B Inverter.

B. Alternate Power Source 72M, via the UPS A Rectifier.

C. Alternate Power Source 72M, via the UPS A Inverter.

D. Normal Power Source 72R, via the 480/120VAC voltage regulator.

Answer: B ILO 2018 NRC Exam Page: 126 of 282 Question 43

Answer Explanation:

The normal power supply to UPS Bus A and the alternate supply to UPS Bus B is from Bus 72M position 3D. The normal supply to UPS Bus B and the alternate supply to UPS Bus A is from Bus 72R position 2B.

During normal operation, each Rectifier/Charger provides regulated DC voltage to the Inverters and maintains the common battery at full charge voltage. In turn, the Inverter supplies its associated distribution panel with AC voltage. If a normal power supply is lost or a Rectifier/Charger fails, the other UPS units Rectifier/Charger will supply both Inverters and maintain the battery charged. If the DC power is lost (both Rectifier/Chargers and the battery), a static switch will automatically supply the distribution cabinet on that side from the alternate power supply.

Since the stem of the question states that UPS B has encountered a rectifier failure, the candidate must recall that UPS has an installed design feature that will allow the other UPS unit's (UPS A in this case) power supply (72M) to take over, via the other unit's rectifier (UPS A) and through the affected unit's (UPS B's) inverter.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly recall the flow of power through the UPS system and assume that failure of the UPS B rectifier will block the input of the normal AND alternate AC sources, which could lead the candidate to determine that power then shifts to the output of the UPS battery through the UPS A inverter. This is not correct because the UPS A rectifier can supply both units' inverters and the UPS battery.

C. The candidate could incorrectly recall the flow of power through the UPS system and assume that the two systems are cross-connected below the two inverters. This could lead the candidate to determine that failure of the UPS B rectifier blocks the flow of power through the UPS B inverter, requiring the UPS A inverter to take over and supply 120Vac power to the UPS B distribution cabinet. This is incorrect because the systems are cross-connected BEFORE the two inverters, allowing UPS A rectifier to feed the UPS B loads through the UPS B inverter and NOT the UPS A inverter.

D. The candidate could incorrectly recall the flow of power through the UPS system and assume that loss of the UPS B rectifier blocks all flow of power through the B UPS unit. This could lead the candidate to determine that the UPS B Static Transfer Switch would transfer to allow 120Vac to the UPS B distribution cabinets from the Alternate Power Source (72M) via the 480VAC/120VAC regulated transformer. This is not correct because the UPS A rectifier can supply both units' inverters and the UPS battery, so the Static Transfer Switch will not have to throw over to the regulated transformer.

Reference Information:

23.308.01, UPS System SOP, Section 1.1 System Description and Enclosure C, UPS System Single Line Diagram.

NUREG 1123 KA Catalog Rev. 2 262002 UPS (AC/DC) 262002 K4. Knowledge of UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) design feature(s) and/or interlocks which provide for the following:

262002 K4.01 3.1/3.4 Transfer from preferred power to alternate power supplies 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK High RO ILO 2018 NRC Exam Page: 127 of 282 Question 43

Associated objective(s):

Uninterruptible Power Supply (R3100)

Cognitive Enabler List the automatic features of Uninterruptible Power Supply System operations.

ILO 2018 NRC Exam Page: 128 of 282 Question 43

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 129 of 282 Question 44

44 K/A Importance: 3.1/3.4 Points: 1.00 R44 Difficulty: 2.00 Level of Knowledge: Low Source: BANK 73986 What would be the result of a loss of Division I ESF DC Power?

A. The Reactor Core Isolation Cooling (RCIC) automatic start function is DISABLED.

B. The High Pressure Coolant Injection (HPCI) automatic start function is DISABLED.

C. The Reactor Water Cleanup (RWCU) Outboard Isolation Valve G3352-F004 will CLOSE.

D. The Inboard Main Steam Isolation Valves (MSIVs) B2103-F022A through D will CLOSE.

Answer: A ILO 2018 NRC Exam Page: 130 of 282 Question 44

Answer Explanation:

Division 1 ESF DC Power supplies power to motor operated valves and RCIC initiation logic. Loss of this power source will DISABLE the automatic start function of the RCIC system, which is a major Division 1 ESF DC Load.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could incorrectly recall that HPCI is a Div 1 System and RCIC is a Div 2 system, which is incorrect. If HPCI were a Div 1 System, then the same impact on RCIC would apply to HPCI, rendering its automatic start function to be DISABLED. Also, two HPCI valves ARE powered by Division 1 DC. One is a steam isolation valve and another isolates the HPCI vacuum relief line. These valves are normally open and would remain open on a Loss of Div 1 ESF DC. If the candidate assumed one or more of these valves went shut on a loss of DC power, then he/she could determine that the HPCI automatic start function was affected.

C. The candidate could incorrectly recall that this valve, or its logic string, is powered from Div 1 ESF DC and therefore assume that it will close upon loss of Div 1 ESF DC. This valve is, in fact, powered from 2PB-1-6B, a Div 2 ESF DC source, and is impacted by RPS B logic, not Div 1 ESF DC, so the valve will stay open.

D. The Inboard MSIVs each have one 3-way DC solenoid that will re-position when they lose power from Division 1 ESF DC. The candidate could determine that, when these 3-way solenoids are de-energized, air to the MSIV actuator will be lost, causing them to close on spring force. However, the DC solenoid valves are in series with 3-way AC solenoids. To close an MSIV, both the DC and the AC solenoids have to de-energize to vent air from the actuator.

Reference Information:

23.309, 260/130VDC Electrical System (ESF and BOP), Enclosure C - ESF DC Electrical Distribution List NUREG 1123 KA Catalog Rev. 2 263000 DC Electrical Distribution 263000 K2. Knowledge of electrical power supplies to the following:

263000 K2.01 Major D.C. loads 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK Low RO Associated objective(s):

DC Electrical Distribution (R3200 & S3102)

Cognitive Enabler Describe how the DC Electrical Distribution System assists in maintaining the critical safety functions.

ILO 2018 NRC Exam Page: 131 of 282 Question 44

45 K/A Importance: 3.4/3.5 Points: 1.00 R45 Difficulty: 3.00 Level of Knowledge: Low Source: BANK 74129 The plant experienced a loss of Bus 64B.

The bus was subsequently restored to the normal lineup EXCEPT the operators neglected to reset the digital load sequencer.

Following the restoration, all power is again lost to Bus 64B.

How will the EDG and electrical distribution system respond to this event?

A. The EDG will automatically start.

The loads on bus 64B will sequence after the output breaker is closed.

B. The EDG will require a manual start.

The loads on bus 64B will sequence after the output breaker is closed.

C. The EDG will require a manual start.

The loads on bus 64B will NOT sequence after the output breaker is closed.

D. The EDG will automatically start.

The loads on bus 64B will NOT sequence after the output breaker is closed.

Answer: A ILO 2018 NRC Exam Page: 132 of 282 Question 45

Answer Explanation:

This questions tests misconceptions on the operation of the EDG Digital Load Sequencer, specifically the operational implication of not resetting the load sequencer and how it will function if it is NOT manually reset.

From Section 3.0 of 23.321: "The associated Digital Load Sequencer must be manually reset upon restoring power to the associated 4160V EDG Bus. Failure to do so will result in the operation of the Digital Load Sequencer following a manual start of the associated EDG and the subsequent closing of its output breaker."

A common misconception is that, if the load sequencer is not reset, the EDG will require manual action to restart. That is not what this precaution says, and not how the load sequencer functions. Contributing to this another misconception regarding where the undervoltage start signal is developed.

The undervoltage start signal comes from the LOAD SHED logic for the respective 4160V ESF bus and NOT from the Load Sequencer. Therefore, if the load sequencer is not reset, the EDG will still automatically start.

Another misconception is that, if the load sequencer is not reset, then it has already actuated and it will not actuate again when the EDG output breaker closes, therefore the loads will not be sequenced back on.

This is also not correct because the load sequencer resets on any subsequent undervoltage (load shed) condition. Therefore, when the EDG output breaker re-closes, the load sequencer will actuate again and re-sequence loads back on without any operator action.

The impact of the precaution is how the load sequencer responds to a MANUAL EDG start (during EDG testing, for example) if it is not reset. If the load sequencer is not reset, and the EDG is manually started, when the EDG output breaker is closed the load sequencer will actuate and go through its normal sequencing, which may be undesirable with the bus energized by the EDG in parallel with offsite power.

Therefore, the correct response of the EDG to this condition is that it will still start, and load properly, if power to a 4160V bus is lost even if the EDG load sequencer was not reset.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could misunderstand the precaution discussed above and believe that the EDG will not automatically start if the load sequencer was not properly reset, which is not the case as explained above.

C. The candidate could misunderstand the precaution discussed above and believe that the EDG will not automatically start if the load sequencer was not properly reset, which is not the case, as explained above. The candidate could also believe that, if not reset, the EDG load sequencer will not actuate when its output breaker closes, so loads will not sequence on as they should, which is also incorrect as explained above.

D. The candidate could correctly recognize that the EDG will automatically start, however, incorrectly determine that, if not reset, the EDG load sequencer will not actuate when its output breaker closes, so loads will not sequence on as they should, which is incorrect as explained above.

Reference Information:

23.321, Engineered Safety Features Auxiliary Electrical Distribution System NUREG 1123 KA Catalog Rev. 2 264000 Emergency Generators (Diesel/Jet) 264000 K5. Knowledge of the operational implications of the following concepts as they apply to EMERGENCY GENERATORS (DIESEL/JET) :

264000 K5.06 3.4/3.5 Load sequencing 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Question Use (ILO 2018)

BANK Low RO ILO 2018 NRC Exam Page: 133 of 282 Question 45

Associated objective(s):

Emergency Diesel Generator (R3000)

Cognitive Enabler Discuss failure modes of Emergency Diesel Generator System controls and vital instruments, including design features that could result in erroneous operation or indication.

Emergency Diesel Generator (R3000)

Cognitive Enabler Discuss failure modes of Emergency Diesel Generator System controls and vital instruments, including design features that could result in erroneous operation or indication.

ILO 2018 NRC Exam Page: 134 of 282 Question 45

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 135 of 282 Question 46

46 K/A Importance: 2.8/2.3 Points: 1.00 R46 Difficulty: 3.00 Level of Knowledge: High Source: NEW 74146 A portion of the Instrument Air System is shown below.

Which of the responses below best completes the following statement to describe the effect on the system if BOTH components highlighted on the drawing were to clog with corrosion products?

Downstream pressure will lower, and when that pressure reaches __________________.

A. 95 psig, the Standby Station Air Compressor will start.

B. 85 psig, the P5000-F401, Station Air to TB Header Isolation Valve will close.

C. 85 psig, the Division 2 Control Air Compressor will start and the P5000-F441, Div 2 Control Air Iso Valve, will be OPEN.

D. 75 psig, the P5000-F441, Div 2 Control Air Iso Valve, will CLOSE AND the Division 2 Control Air Compressor will then automatically start.

Answer: C ILO 2018 NRC Exam Page: 136 of 282 Question 46

Answer Explanation:

The components highlighted are filters in the normal air supply line from the Interruptable Air System (IAS) to the Non-Interruptable Air System (NIAS).

If these filters were to clog, air flow downstream (to the right) of the filters would lower, resulting in lowering pressure sensed by pressure switch PSE-N482B.

This pressure switch causes an auto start of the Division 2 Control Air Compressor (CAC) at 85 psig (see ARP 7D50, Control Air Compressor Auto Start). At this pressure, the P5000-F441, Div 2 Control Air Iso Valve, will be OPEN, since it does not get a signal to close until pressure sensed by PSE-N482B reaches 75 psig (See Student Text ST-OP-315-0071, Table 7).

Therefore, when pressure sensed by that switch reaches 85 psig, the CAC will auto start and the P5000-F441 will still be OPEN.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The Standby Station Air Compressors do receive a signal to start automatically at 95 psig. This is incorrect, however, because the pressure switch that causes this action (P5000-N415) is located upstream of the filters (not shown on the drawing in the stem of the question) and would not be impacted by clogging of the filters highlighted in the question.

B. The P5000-F401 does receive a signal to isolate automatically at 85 psig. This is incorrect, however, because the pressure switch that causes this action (P5000-N404) is located upstream of the filters (not shown on the drawing in the stem of the question) and would not be impacted by clogging of the filters highlighted in the question.

D. The candidate could correctly recall that the P5000-F441 valve will close at 75 psig. However, the candidate could incorrectly determine that the Division 2 CAC starts at the same time the Div 2 isolation valve closes, at 75 psig, which is not correct since the CAC gets a signal to start at 85 psig as sensed by PSE-N482B.

Reference Information:

ST-OP-315-0071, Compressed Air System Student Text.

M-5730-3, NIAS Division I & II FOS.

7D50 Control Air Compressor Auto Start.

NUREG 1123 KA Catalog Rev. 2 300000 Instrument Air System 300000 K6. Knowledge of the effect that a loss or malfunction of the following will have on the INSTRUMENT AIR SYSTEM:

300000 K6.13 2.8/2.3 Filters 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Compressed Air Systems (P5001 & P5002)

Cognitive Enabler List the automatic features of Compressed Air System operations.

Compressed Air Systems (P5001 & P5002)

Cognitive Enabler List the automatic features of Compressed Air System operations.

ILO 2018 NRC Exam Page: 137 of 282 Question 46

ILO 2018 NRC Exam Page: 138 of 282 Question 46 THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 139 of 282 Question 47

47 K/A Importance: 2.7/2.7 Points: 1.00 R47 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 74166 The plant is at 100% power. The crew is preparing to alternate running RBCCW pumps, which will require them to start the South and then shut down the North.

Which of the following describes the response that the CRLNO will observe when the South RBCCW Pump is started and why?

A. The P42-F403, RBCCW DP Control Valve, will OPEN to maintain a constant DP between the RBCCW and GSW headers.

B. The P42-F403, RBCCW DP Control Valve, will CLOSE to maintain a constant DP between the RBCCW and GSW headers.

C. The P42-F403, RBCCW DP Control Valve, will OPEN to maintain a constant DP between the RBCCW supply and return headers.

D. The P42-F403, RBCCW DP Control Valve, will CLOSE to maintain a constant DP between the RBCCW supply and return headers.

Answer: C ILO 2018 NRC Exam Page: 140 of 282 Question 47

Answer Explanation:

Per a NOTE in section 6.1 of 23.127, when the standby (South) RBCCW Pump is started, the P42-F403 RBCCW DP Control Valve, will OPEN to compensate for the increased flow and pressure.

This recirculation control valve maintains a constant pressure differential across the supply and return headers, based on RBCCW Pump d/p. By doing this, the flowrate through the various loads will not change thereby minimizing temperature fluctuations on RBCCW cooled components.

The candidate must recognize that, when the system controls are manipulated to start the standby RBCCW Pump, then RBCCW Supply Pressure would tend to rise. However, the candidate must then determine that the P42-F403 is designed to counteract this rise in pressure and will OPEN in response to this as it attempts to maintain the d/p between the RBCCW supply and return headers constant.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could correctly recall that the P42-F403 will OPEN in this situation, but incorrectly recall that the valve maintains d/p between RBCCW and GSW, which it does not.

B. The candidate could incorrectly recall that the P42-F403 responds to maintain d/p between the RBCCW and GSW headers and conclude that, if an additional pump is started, this valve would have to CLOSE in order to maintain this differential pressure. This is plausible because other systems at Fermi 2 (Main Turbine Lube Oil, Hydrogen Seal Oil) have pressure control valves that would CLOSE, in response to an additional pump start, in an attempt to maintain system pressure or differential pressure.

D. The candidate could correctly recall that the P42-F403 maintains the d/p between the RBCCW supply and return headers. However, the candidate could incorrectly determine that this valve would CLOSE in response to starting the standby RBCCW Pump. This is plausible because other systems at Fermi 2 (Main Turbine Lube Oil, Hydrogen Seal Oil) have pressure control valves that would CLOSE, in response to an additional pump start, in an attempt to maintain system pressure or differential pressure.

Reference Information:

23.127, RBCCW/EECW System SOP, Section 6.1 Alternating RBCCW Pumps.

ST-OP-315-0067, RBCCW/EECW System Student Text.

NUREG 1123 KA Catalog Rev. 2 400000 Component Cooling Water System 400000 A1. Ability to predict and / or monitor changes in parameters associated with operating the CCWS controls including:

400000 A1.03 2.7/2.7 CCW Pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Question Use (ILO 2018)

Low NEW RO ILO 2018 NRC Exam Page: 141 of 282 Question 47

Associated objective(s):

Reactor Building Closed Cooling Water Emergency Equipment Cooling Water (P4200/P4400)

Cognitive Enabler List the automatic features of Reactor Building Closed Cooling Water/Emergency Equipment Cooling Water System operations.

Reactor Building Closed Cooling Water Emergency Equipment Cooling Water (P4200/P4400)

Cognitive Enabler List the automatic features of Reactor Building Closed Cooling Water/Emergency Equipment Cooling Water System operations.

ILO 2018 NRC Exam Page: 142 of 282 Question 47

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 143 of 282 Question 48

48 K/A Importance: 3.2/3.3 Points: 1.00 R48 Difficulty: 3.00 Level of Knowledge: High Source: NEW 74206 A LOCA has occurred. RHR Pump D is operating in the LPCI Mode, at 10,000 gpm, maintaining RPV Level steady at 197 inches.

2D26, DIV II RHR SYSTEM LOW FLOW BYPASS INITIATED alarms due to a flow instrument failure and the RHR system responds as if this were a valid condition.

Which one of the following describes the impact of this failure and the action that should be taken to mitigate its consequences?

A. RHR Pump D will trip due to suction valve closure; START RHR Pump B.

B. RHR Pump D will be operating with NO discharge flow path; STOP RHR Pump D.

C. RPV level will rise due to water bypassing the RHR Heat Exchanger; CLOSE E1150-F048B, Div 2 RHR HX Bypass Valve.

D. RPV level will lower due to water being pumped to the Torus; CLOSE the E1150-F007B, Div 2 RHR Pumps Minimum Flow Valve.

Answer: D ILO 2018 NRC Exam Page: 144 of 282 Question 48

Answer Explanation:

Alarm 2D26 comes in when instrument E1100-N021B senses low flow (<3,000 gpm) through the RHR system. The impact of this will be an automatic opening of the E1150-F007B, Div 2 RHR Pumps Min Flow Valve. When open, this valve will pass 3,000 gpm of water directly from the RHR Pump discharge back to the Torus, which will result in a lowering RPV Water Level. The correct action, per ARP 2D26, is to attempt to CLOSE the E1150-F007B.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could assume that the alarm was the result of a low flow condition or the alarm could result in an automatic closure of the RHR Suction Valve(s). If the suction valve did close, this would result in a trip of the running RHR Pump, so starting another RHR Pump to maintain RPV level would be the correct course of action. This is incorrect, however, because the alarm results in opening of the Minimum Flow Bypass Valve.

B. The candidate could assume that the alarm will result in automatic closure of a valve in the RHR Pump discharge path, either directly or upon interlock from opening of the Minimum Flow Valve. Also, if an RHR Pump were operating with no discharge flow path, alarm 2D26 would occur, which also makes this distractor plausible. The operator could determine the correct course of action is to stop the running RHR Pump under this condition. This distractor is incorrect, however, because the alarm results in opening of the Minimum Flow Bypass Valve.

C. The candidate could assume that the alarm resulted in opening of the Heat Exchanger BYPASS and not the Minimum Flow BYPASS valve, thereby causing RPV temperature and level to rise, due to thermal expansion of the water during heatup, and also reduced headloss. The candidate could then determine that closing the HX Bypass Valve is the correct course of action to take. This answer is incorrect, however, because the alarm will cause the Minimum Flow Bypass Valve to open, thereby lowering RPV water level.

Reference Information:

2D26, Div II RHR System Low Flow Bypass Initiated.

NUREG 1123 KA Catalog Rev. 2 203000 RHR/LPCI: Injection Mode 203000 A2. Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

203000 A2.13 3.2/3.3 Valve openings.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

High NEW RO ILO 2018 NRC Exam Page: 145 of 282 Question 48

Associated objective(s):

Residual Heat Removal (E1100)

Cognitive Enabler Describe general RHR System operation, including component operating sequence, normal operating parameters, and expected system response.

Residual Heat Removal (E1100)

Cognitive Enabler Describe general RHR System operation, including component operating sequence, normal operating parameters, and expected system response.

ILO 2018 NRC Exam Page: 146 of 282 Question 48

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 147 of 282 Question 49

49 K/A Importance: 3.2/3.2 Points: 1.00 R49 Difficulty: 3.00 Level of Knowledge: High Source: NEW 74207 The Plant is in Cold Shutdown. RHR pump B is aligned and operating in the Shutdown Cooling Mode. A rise in RPV pressure is detected as follows:

• B31-N111A, Division 1 Reactor Vessel Shutdown Cooling Cut-in Permissive Interlock Pressure Instrument - 95 psig

• B31-N111B, Division 2 Reactor Vessel Shutdown Cooling Cut-in Permissive Interlock Pressure Instrument - 85 psig You observe the following RHR System conditions:

• E1150-F009, RHR SDC Inbd Suction Iso Valve is CLOSED.

• E1150-F008, RHR SDC Otbd Suction Iso Valve is OPEN.

• E1150-F004B, RHR Pump B Torus Suct Iso Valve is CLOSED.

• E1150-F006B, RHR Pump B SDC Suct Iso Valve is OPEN.

• E1102-C002B, Div 2 RHR Pump B is RUNNING.

Which of the following automatic operations FAILED to occur and requires manual operator action to correct?

A. E1150-F009, failed to OPEN.

B. E1102-C002B failed to TRIP.

C. E1150-F008, failed to CLOSE.

D. E1150-F006B, failed to CLOSE.

Answer: B ILO 2018 NRC Exam Page: 148 of 282 Question 49

Answer Explanation:

The Reactor Vessel Shutdown Cooling Cut-In Permissive Interlock setpoint is 89.5 psig. There are two instruments that sense this pressure to enforce this interlock, they are the B31-N111A and B31-N111B.

The logic is such that the A instrument supplies Division 1 and the B instrument Division 2. If the A instrument senses pressure above 89.5 psig, then the E1150-F009 will close. Conversely, if the B instrument senses pressure above 89.5 psig, then the E1150-F008 will close. Therefore, the valves in the stem of the question have responded correctly.

The RHR Pumps have interlocks that cause them to trip on loss of suction path from either the RPV or the Torus. An RHR Pump will trip if either its Torus Suction path (E1150-F004 not fully open) AND its SDC Suction path (F009, F008 and/or F006 are not fully open) is not aligned. With the F009 and the F004B closed, the candidate should recognize a condition that would require an automatic pump trip.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly recall the SDC Cut-In Permissive is met when pressure is ABOVE the setpoint, and not below, and therefore conclude that the Division 1 (Inboard) permissive is met, so the E1150-F009 should be open, which is incorrect as described above.

C. The candidate could incorrectly recall the logic for automatic closure of the RHR SDC Suction Valves on an overpressure condition and conclude that the one instrument above the setpoint should have caused both valves to close, which is incorrect per the logic described above.

D. The candidate could incorrectly determine that the combination of high pressure on the Division 1 instrument, along with closure of the F009, would result in closure of the SDC Pump Suction valve, which is incorrect as described above.

Reference Information:

23.601, Instrument Trip Sheets ST-OP-315-0041, RHR System Student Text, Table 4 Control Functions and Interlocks.

NUREG 1123 KA Catalog Rev. 2 205000 RHR Shutdown Cooling Mode 205000 A3. Ability to monitor automatic operations of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) including:

205000 A3.02 3.2/3.2 Pump trips 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Residual Heat Removal (E1100)

Cognitive Enabler List the interlocks associated with RHR System components.

Residual Heat Removal (E1100)

Cognitive Enabler List the interlocks associated with RHR System components.

ILO 2018 NRC Exam Page: 149 of 282 Question 49

50 K/A Importance: 4.0/3.9 Points: 1.00 R50 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 74227 The plant was operating at 100% power when an inadvertent HPCI initiation occurred. The CRLNO has evaluated plant conditions and determined that HPCI injection is not needed.

Which of the following contains the actions that are needed to trip the HPCI Turbine and keep it in a tripped condition?

A. Turn the arming collar to ARMED, depress then release the TRIP Pushbutton.

B. Turn the arming collar to ARMED, depress the TRIP Pushbutton, then release the TRIP Pushbutton when turbine speed indicates zero rpm.

C. Turn the arming collar to ARMED, depress the TRIP Pushbutton, place the Aux Oil Pump in OFF when turbine speed indicates zero rpm, then release the TRIP Pushbutton.

D. Turn the arming collar to ARMED, depress the TRIP Pushbutton, place the Aux Oil Pump in OFF when turbine speed indicates zero rpm, depress the HPCI Initiation Signal RESET pushbutton, then release the TRIP Pushbutton.

Answer: C ILO 2018 NRC Exam Page: 150 of 282 Question 50

Answer Explanation:

Per 23.202, Section 8.1, the actions listed are required to trip the HPCI Turbine. To summarize:

• The Operator must ARM the Trip Pushbutton.

• The Operator must depress and HOLD the Trip Pushbutton.

• The Operator will observe various HPCI System indications to verify proper system shutdown.

• When Turbine RPM reaches 0, the Operator must place the Aux Oil Pump in OFF to prevent HPCI from re-starting.

• Then, the Operator can release the Trip Pushbutton.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could confuse the HPCI system with the RCIC system and determine that all that is required is to arm, then depress the HPCI Trip Pushbutton. For RCIC, this will unlatch the Trip Throttle Valve, which will keep the system in a tripped condition (see 23.206, Section 8.1 RCIC Shutdown). For HPCI, if the Trip Pushbutton is release, HPCI will restart. It is plausible, but incorrect, for the student to assume that the HPCI Trip controls work the same as RCIC due to the similarity between the systems.

B. The candidate could recall that the Trip Pushbutton needs to be armed and depressed and recall that the operator must verify that HPCI stops rotating by observing turbine RPM. The candidate could then incorrectly determine that, once turbine RPM is zero, oil pressure developed by the installed turbine driven oil pump is too low to support turbine restart, which is true. However, the candidate would fail to recall that HPCI has a motor driven oil pump that will start on low oil pressure, when the turbine driven oil pump speed is too low, and the motor driven oil pump develops sufficient pressure to support turbine restart.

D. The candidate could determine that the initiation signal must be reset to prevent HPCI from restarting, which is not correct. Further down the actions of section 8.1, resetting the initiation signal is only necessary if the initiating condition is clear and before restoring HPCI to standby. The stem of the question asked what it takes to TRIP the HPCI system and NOT restore it to standby.

Reference Information:

23.202, HPCI System SOP, Section 8.1 HPCI Shutdown.

23.206, RCIC System SOP, Section 8.1 RCIC Shutdown.

NUREG 1123 KA Catalog Rev. 2 206000 HPCI System.

206000 A4. Ability to manually operate and/or monitor in the control room:

206000 A4.12 4/3.9 Turbine trip controls: BWR-2,3,4 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

High Pressure Coolant Injection Cognitive Enabler Identify associated remote and local instrumentation, indications, alarms, and controls for the HPCI System.

High Pressure Coolant Injection Cognitive Enabler Identify associated remote and local instrumentation, indications, alarms, and controls for the HPCI System.

ILO 2018 NRC Exam Page: 151 of 282 Question 50

ILO 2018 NRC Exam Page: 152 of 282 Question 50 THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 153 of 282 Question 51

51 K/A Importance: 3.7/3.9 Points: 1.00 R51 Difficulty: 3.00 Level of Knowledge: High Source: NEW 74228 Below is a numbered list of instruments associated with monitoring the status of the Core Spray System when it is aligned to the Torus:

1. G51-R402, Narrow Range Torus Level Indicator

2. T50-R804A(B), Div 1(2) Torus Level Recorder

3. T23-R800, Torus Water Temperature Recorder

4. T50-R800A(B), Div 1(2) PC Air and Water Temperatures Recorder The plant is in a lowering Torus Water Level event in which Core Spray is needed to maintain RPV Water Level.

With Torus Level below -11", which of the above are the Post-Accident Instruments that CAN BE USED to verify Core Spray is within its Vortex and NPSH Limits?

A. ONLY 2 and 4 can be used.

B. Either 1 OR 2 can be used WITH 4.

C. ONLY 2 can be used with either 3 OR 4.

D. ALL of the above instruments can be used.

Answer: A ILO 2018 NRC Exam Page: 154 of 282 Question 51

Answer Explanation:

The candidate must first recall that the Core Spray Vortex and NPSH Limits are impacted by a combination of Torus Water Level, and Torus Water Temperature, among others. The candidate must then determine that instrument 1 is NOT a post-accident instrument and, for the conditions given in the stem of the question, cannot be used to verify CS is within NPSH and Vortex limits.

Also, with Torus Water Level below -11", the candidate must recall that instrument 3 is unreliable (per Caution 6 of the EOPs) so, although it IS a post-accident instrument, it cannot be used under these conditions to verify CS is within limits.

Therefore, the candidate must conclude that only instruments 2 and 4 are post-accident instruments that can be used under the conditions given in the stem of the question.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could incorrectly determine that instrument 1 is a post-accident instrument that can be used under these conditions, which is incorrect as described above.

C. The candidate could incorrectly determine that both instruments 3 and 4 are post-accident instruments (which is true) that can be used to determine CS system status. However, instrument 3 cannot be used with Torus Water Level below -11" per Caution 6 of the EOPs.

D. The candidate could incorrectly recall that all instruments listed are post-accident instruments that can be used to determine CS system status under the conditions given in the stem of the question, which is incorrect as described above.

Reference Information:

29.100.01, EOP Curves, Cautions and Tables.

NUREG 1123 KA Catalog Rev. 2 209001 Low Pressure Core Spray System.

G2.4.3 3.7/3.9 Ability to identify post-accident instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Core Spray System Cognitive Enabler Discuss effective monitoring of the Core Spray System using local, remote, computer displays and alarms.

Core Spray System Cognitive Enabler Discuss effective monitoring of the Core Spray System using local, remote, computer displays and alarms.

ILO 2018 NRC Exam Page: 155 of 282 Question 51

52 K/A Importance: 2.7/2.7 Points: 1.00 R52 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 74248 Which of the following instruments would be adversely affected by a break in the SLC injection piping within the Reactor Pressure Vessel?

A. CRD Drive Water Pressure.

B. Core Spray Discharge Pressure.

C. Core Plate Differential Pressure.

D. Calibrated Jet Pump Differential Pressure.

Answer: C ILO 2018 NRC Exam Page: 156 of 282 Question 52

Answer Explanation:

Note: Refer to M-5701-2, Nuclear Boiler System FOS, Grid D5 for this explanation.

The SLC system penetration and piping provides above core plate pressure for the Core Spray line break detection instrumentation from the outside pipe. The same penetration is also used for Core Plate DP, Jet Pump DP: and CRD Drive Water DP, from the inside pipe (SLC discharge sparger pipe).

A break in the SLC injection piping would adversely affect the Core Plate Differential Pressure instrument, which is the correct answer for this question.

This break would also adversely affect the CRD Drive Water Differential Pressure, Core Spray Break Detection and non-calibrated Jet Pump Differential Pressure instruments.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly recall that the SLC penetration impacts the CRD Drive Water Pressure instrument, which is incorrect but plausible because this SLC penetration provides the RPV Pressure (above Core Plate pressure) input to the CRD Drive Water Differential Pressure instrument and not the Drive Water Pressure instrument.

B. The candidate could incorrectly recall that the SLC penetration impacts the Core Spray Discharge Pressure instrument, which is incorrect but plausible because this SLC penetration provides the RPV Pressure (above Core Plate pressure) input to the Core Spray Break Detection (differential pressure) instrument and not the Core Spray Discharge Pressure instrument.

D. The candidate could incorrectly recall that the SLC penetration impacts the Calibrated Jet Pump Differential Pressure instrument, which is incorrect but plausible because this SLC penetration provides the high-pressure input, i.e. diffuser pressure, from the below Core Plate pressure tap to the NON calibrated Jet Pumps. The Calibrated Jet Pumps (Jet Pumps 5, 10, 15 & 20) have a separate tap directly in their diffuser section that provides the high-pressure input to the Calibrated Jet Pump Differential Pressure instruments.

Reference Information:

ST-OP-315-0014 SLC System Student Text, System Interrelations section on the Core Spray System.

M-5701-2 Nuc Boiler Instrumentation FOS.

NUREG 1123 KA Catalog Rev. 2 211000 SLC System 211000 K1. Knowledge of the physical connections and/or cause-effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following:

211000 K1.02 2.7/2.7 Core plate differential pressure indication 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (2) General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Standby Liquid Control Cognitive Enabler Discuss the Standby Liquid Control System interrelationships with other systems.

Standby Liquid Control Cognitive Enabler Discuss the Standby Liquid Control System interrelationships with other systems.

ILO 2018 NRC Exam Page: 157 of 282 Question 52

53 K/A Importance: 3.5/3.7 Points: 1.00 R53 Difficulty: 3.00 Level of Knowledge: Low Source: NEW 74269 The plant is operating at 96% Power.

A Pressure Control System malfunction results in indicated RPV pressure rising to 1100 psig.

During the transient, the B21-N678A, Reactor Vessel Steam Dome Pressure - High Trip Unit, remains UNTRIPPED.

How will the Reactor Protection System (RPS) be affected by the above conditions?

A. BOTH RPS Trip Systems will TRIP.

B. ONLY RPS Trip System A will TRIP.

C. ONLY RPS Trip System B will TRIP.

D. NEITHER RPS Trip System will TRIP.

Answer: A ILO 2018 NRC Exam Page: 158 of 282 Question 53

Answer Explanation:

NOTE: Refer to 23.601, Page 10 for this discussion. Trip Logic for the Reactor Vessel Steam Dome Pressure - High Trip Function for RPS is as follows:

Trip System A = Trip Channels A (A1), C (A2)

Trip System B = Trip Channels B (B1), D (B2)

Trip Channel A1 = B21-N678A; A2 = B21-N678C Trip Channel B1=B21-N678B; B2 = B21-N678D The Trip Setpoint for this Function is </= 1093 psig.

Trip Logic for this Function is (A1 or A2) and (B1 or B2) = Full Rx Scram With the failure given in the stem of the question, the B, C and D instruments will trip, resulting in Trip Channels A2, B1 and B2 tripping, which will result in BOTH RPS Trip Systems A and B tripping, which will result in a FULL reactor scram.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could incorrectly recall RPS Trip Logic and determine the instrument failure will prevent the B RPS Trip System from tripping, which is incorrect as can be seen from the Trip Logic described above.

C. The candidate could incorrectly recall RPS Trip Logic and determine the instrument failure will prevent the A RPS Trip System from tripping, which is incorrect as can be seen from the Trip Logic described above.

D. The candidate could incorrectly recall RPS Trip Logic and assume that all four pressure instrument must trip and therefore determine that the instrument failure will prevent BOTH RPS Trip Systems from tripping, which is incorrect as can be seen from the Trip Logic described above.

Reference Information:

23.601 , Instrument Trip Sheets.

NUREG 1123 KA Catalog Rev. 2 212000 RPS 212000 K6. Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR PROTECTION SYSTEM:

212000 K6.03 3.5/3.7 Nuclear boiler instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Reactor Protection System (C7100)

Cognitive Enabler Describe general Reactor Protection System operation, including component operating sequence, normal operating parameters, and expected system response.

Reactor Protection System (C7100)

Cognitive Enabler Describe general Reactor Protection System operation, including component operating sequence, normal operating parameters, and expected system response.

ILO 2018 NRC Exam Page: 159 of 282 Question 53

ILO 2018 NRC Exam Page: 160 of 282 Question 53 THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 161 of 282 Question 54

54 K/A Importance: 2.8 Points: 1.00 R54 Difficulty: 3.00 Level of Knowledge: Low Source: NEW 73327 NOTE: The drawing above is shown with all lights OFF (not lit).

When the switch shown is rotated to the TEST position, what occurs to allow for testing of the Scram Discharge Volume isolation valves?

A. Air is isolated and vented to test closure of the valves.

B. Air is isolated and vented to test opening of the valves.

C. Air is supplied to the actuators to test closure of the valves.

D. Air is supplied to the actuators to test opening of the valves.

Answer: A ILO 2018 NRC Exam Page: 162 of 282 Question 54

Answer Explanation:

The candidate must recognize that the design feature that has been installed to allow testing of the Scram Discharge Volume (SDV) isolation valves is a keylock switch on the H11-P603 panel that, when taken to TEST, energizes a valve which isolates and vents air from the SDV vent and drain valves to test their closing ability.

Distractor Explanation:

B. Distractor Is incorrect because the valves are normally open and rotating the keylock switch to TEST isolates and vents air to close them, not open them. This distractor is plausible if the candidates incorrectly remembered that the keylock switch tested the valves in the open direction with a scram signal in, rather than how we actually test them, which is in the close direction. This distractor is also plausible if the candidate incorrectly recalled the normal lineup of the valves.

C. Distractor is incorrect because rotating the keylock switch to TEST isolates and vents air to close them, it does not supply air to the actuator to close them. This distractor is plausible because some valves in the Scram Air Header close when air is supplied to their actuators and some open when air is supplied to their actuators. If a candidate incorrectly thought that these valves need air to close, then this distractor could be selected.

D. Distractor is incorrect because the valves are normally open and rotating the keylock switch to TEST isolates and vents air to close them, not open them. Distractor is plausible if the candidate incorrectly remembered that the keylock switch tested the valves in the open direction with a scram signal in and if the candidate thought that air was needed to open the valves. This distractor is also plausible if the candidate incorrectly recalled the normal lineup of the valves.

Reference Information:

M-5703-2, CRD Scram Discharge Volume Functional Operating Sketch (FOS).

Student Text ST-OP-315-0010 NUREG 1123 KA Catalog Rev. 2 201001 K4. Knowledge of CONTROL ROD DRIVE HYDRAULIC SYSTEM design feature(s) and/or interlocks which provide for the following: Testing SCRAM discharge volume isolation valves.

201001 K4.07 Testing SCRAM discharge volume isolation valves 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Control Rod Drive Hydraulics (C1150)

Cognitive Enabler Identify associated remote and local instrumentation, indications, alarms, and controls for the Control Rod Drive Hydraulic system.

ILO 2018 NRC Exam Page: 163 of 282 Question 54

55 K/A Importance: 4.1 Points: 1.00 R55 Difficulty: 3.00 Level of Knowledge: High Source: NEW 73346 The plant is operating at 100% power when a reactor scram occurs.

What is the (1) impact of the scram on Control Rod Drive Mechanism (CRDM) temperatures and (2) action that can be taken to mitigate this impact?

A. (1) They will rise (2) Reset the reactor scram.

B. (1) They will lower (2) Reset the reactor scram.

C. (1) They will rise (2) Open the CRD Flow Control Valve D. (1) They will lower (2) Close the CRD Flow Control Valve.

Answer: A ILO 2018 NRC Exam Page: 164 of 282 Question 55

Answer Explanation:

Following a scram, after the CRD accumulators discharge, the scram stroke is completed by RPV pressure unseating the internal ball check valve of the CRDM, thereby allowing reactor water into the under-piston volume. Reactor water temperature overcomes the cooling effect of CRDH cooling water flow (nominally 63 gpm spread out over 185 Control Rod Drive Mechanisms) and results in rising CRDM internal temperatures. Furthermore, the normal scram response of the CRD FCV is to CLOSE, which lowers cooling water flow even more (to about 15 gpm spread out over 185 CRDMs).

ARP 3D13, CRD Hydraulic Temperature High, has a note which states "This alarm will nearly always accompany a Reactor Scram. Resetting the scram should result in lowering temperatures".

Therefore, a reactor scram causes CRDM temperatures to rise, which is corrected by resetting the scram.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could incorrectly conclude that the HCU Accumulator discharge, caused by the Scram action, will cool the CRDMs. However, this is incorrect because CRDM temperatures will rise as evidenced by the note in ARP 3D13. Resetting the scram is the correct action to take but it is taken to correct a rising temperature condition (per the note in ARP 3D13) and not a lowering temperature condition.

C. This distractor is incorrect because, even though CRDM temperatures will rise, there is no procedural guidance to perform open the CRD FCV under these conditions. To the contrary, AOP 20.106.03, CRD FCV Failure, has a note which directs CLOSING the CRD FCV on a scram condition, with the FCV in manual, to allow for limiting pump flow due to runout considerations and to allow for faster recharging of the scram accumulators. Since the normal scram action of this valve is to close, opening it to correct the rising temperature condition is not warranted.

D. The candidate could incorrectly conclude that the HCU Accumulator discharge, caused by the Scram action, will cool the CRDMs so closing the CRD FCV is the correct action. However, this is incorrect because CRDM temperatures will rise as evidenced by the note in ARP 3D13. Also, the action to close the CRD FCV is not necessary because the valve's normal scram function is to close automatically.

Reference Information:

ARP 3D13, CRD Hydraulic Temperature High AOP 20.106.03, CRD Flow Control Valve Failure Student Text ST-OP-315-0009 NUREG 1123 KA Catalog Rev. 2 201003 A2. Ability to (a) predict the impacts of the following on the CONTROL ROD AND DRIVE MECHANISM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

201003 A2.05 4.1*/4.1 Reactor Scram 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Question Use (ILO 2018)

High NEW RO ILO 2018 NRC Exam Page: 165 of 282 Question 55

Associated objective(s):

Control Rod Drive Mechanism (B1104, C1101 & C1102)

Cognitive Enabler Describe general Control Rod Drive Mechanism operation, including component operating sequence, normal operating parameters, and expected system response.

Control Rod Drive Mechanism (B1104, C1101 & C1102)

Cognitive Enabler Describe general Control Rod Drive Mechanism operation, including component operating sequence, normal operating parameters, and expected system response.

ILO 2018 NRC Exam Page: 166 of 282 Question 55

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 167 of 282 Question 56

56 K/A Importance: 2.9 Points: 1.00 R56 Difficulty: 3.00 Level of Knowledge: Low Source: NEW 73366 See the attached Rod Pull Sheet to answer this question.

A startup is in progress with count rate approaching criticality. What is the operational implication of the P603 Operator selecting control rod 26-39 for movement?

A. A Select Error (SE) would be generated. ONLY withdrawal of this control rod would be prevented.

B. A Select Error (SE) would be generated. ANY movement of this control rod would be prevented.

C. No error would be generated. Rod movement could occur in either direction at which time a rod block, due to Insert Error (IE) or Withdrawal Error (WE), would occur.

D. A Select Error (SE) would be generated. Rod movement could occur in either direction at which time a rod block, due to Insert Error (IE) or Withdrawal Error (WE), would occur.

Answer: B ILO 2018 NRC Exam Page: 168 of 282 Question 56

Answer Explanation:

RWM sequence enforcement restricts movement of control rods that are not selected and moved in compliance with the selected sequence below the LPSP. Rod motion that would result in an insert or withdraw error is not permitted. Sequence enforcement is active between all rods full-in and the LPSP.

Select Errors (SE) will be indicated upon selection of any rod which is not in sequence order. Rod movement must not only conform to the RWM step, but also the rod order within the step.

Since the selected control rod is NOT the next rod in the sequence, a Select Error will occur that results in an insert block (IB) and withdrawal block (WB). Therefore, movement of this control rod, in either direction, will be blocked due to selecting this control rod out of sequence.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could recognize that the selected rod is not the next one in the sequence, resulting in a Select Error. However, the candidate could incorrectly assume that all rods in a group (on a page of the pull sheets) can move inward to their group insertion limit but the rod could not move outward until the previous rods in the group were at their group withdrawal limit. This is plausible because all the rods in a group are on the same RWM step, but it is incorrect because rod movement must not only conform to the RWM step but also the rod order within the step.

C. The candidate could fail to recognize that selecting a rod not in sequence will result in a Select Error.

This could lead the candidate to determine that the rod is free to move out of its current position and, when it does, the Rod Worth Minimizer would THEN impose either rod block (due to either IE or WE, depending on direction of rod travel).

D. The candidate could recognize that the selected rod is not the next one in the sequence, resulting in a Select Error. The candidate could fail to recognize that a Select Error (SE) results in both a Insert Block (IB) and Withdrawal Block (WB). This could lead the candidate to determine that the rod could move out of its current position and, as soon as it does, the resulting Insert Error (IE) or Withdrawal Error (WE) is what would THEN generate the rod block (due to either IB or WB, depending on direction of rod travel). This is incorrect because the SE will generate both an IB and WB as described above.

Reference Information:

SOP 23.608, Rod Worth Minimizer System Student Text ST-OP-315-0013 NUREG 1123 KA Catalog Rev. 2 201006 K5. Knowledge of the operational implications of the following concepts as they apply to ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC):

201006 K5.08 2.9/2.9 Operating sequence: P-Spec(Not-BWR6) 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Rod Worth Minimizer (C1108)

Cognitive Enabler List the interlocks associated with Rod Worth Minimizer system components.

ILO 2018 NRC Exam Page: 169 of 282 Question 56

57 K/A Importance: 3.4/3.5 Points: 1.00 R57V2 Difficulty: 3.00 Level of Knowledge: Low Source: BANK 73386 The Plant is operating at 100% power with both RWCU Pumps and Filter/Demineralizers in service.

Alarm 2D110, RWCU NON REGEN HX OUTLET TEMP HIGH, annunciates due to a sensed temperature of 142°F.

In what way, if any, will the RWCU system automatically respond?

A. No automatic system response will occur.

B. G3352-F119, RWCU Supply Suct Iso Vlv, will shut and the RWCU Pumps will trip on low flow.

C. G3352-F119, RWCU Supply Suct Iso Vlv, will shut and the RWCU Pumps will remain running.

D. G3352-F001, RWCU Inboard Containment Iso Vlv, G3352-F004, RWCU Outboard Containment Iso Vlv, and G3352-F220, RWCU Return Iso Valve will shut and the RWCU Pumps will trip on interlock.

Answer: B ILO 2018 NRC Exam Page: 170 of 282 Question 57

Answer Explanation:

The G3352-F119 will close on the sensed high temperature, which will result in an indirect trip of the RWCU Pumps on Low Flow.

The NRHX Outlet Temperature High Isolation has no accident mitigation function. The basis for this isolation is to protect the ion exchange resin from deterioration due to high temperature. The low flow trip protects the pumps from damage due to running on low flow.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could determine that the temperature given is not valid for the alarm condition and determine that automatic closure of the G3352-F119 would not occur. Or, the candidate could recognize the alarm and temperature as being valid, but not correctly recall the automatic system response that would take place as described above. Either of these reasons could lead the candidate to determine that no automatic system response will occur, which is not correct as described above.

C. The candidate could recognize the initiating condition as a signal that causes the G3352-F119 to close. However, the candidate could fail to recognize that closure of the G3352-F119 will result in a low flow condition in the system that will in turn cause the RWCU pumps to trip.

D. The candidate could incorrectly recall the high temperature trip as an isolation signal and conclude that the RWCU system isolation valves (G3352-F001, F004 and F220) would close and cause the RWCU pumps to immediately trip, since the RWCU pumps do immediately trip (direct trip) on closure of any of these valves.

Reference Information:

ARP 2D110, RWCU Non-Regen HX Outlet Temp High 23.707, RWCU System SOP Student Text ST-OP-315-0008 NUREG 1123 KA Catalog Rev. 2 204000 A3. Ability to monitor automatic operations of the REACTOR WATER CLEANUP SYSTEM including:

204000 A3.04 3.4/3.5 Response to interlocks and trips designed to protect system components 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK Low RO Associated objective(s):

Reactor Water Cleanup Cognitive Enabler Discuss effective monitoring of the Reactor Water Cleanup system using local, remote, computer displays and alarms.

Reactor Water Cleanup Cognitive Enabler Discuss effective monitoring of the Reactor Water Cleanup system using local, remote, computer displays and alarms.

ILO 2018 NRC Exam Page: 171 of 282 Question 57

58 K/A Importance: 3.1/3.4 Points: 1.00 R58 Difficulty: 3.00 Level of Knowledge: Low Source: NEW 73406 The plant is operating in the Power Range with a Traversing In-Core Probe (TIP) run in progress.

A LOCA occurs resulting in a reactor scram on high Drywell pressure.

A malfunction of the Primary Containment Isolation System (PCIS) results in Group 15 -

Traversing In-core Probe System, NOT receiving an isolation signal.

What is the effect of this malfunction on the TIP system?

A. The TIP detector WILL automatically withdraw, however the Ball Valve WILL NOT automatically close. The Ball Valve can be manually actuated to establish containment integrity.

B. The TIP detector WILL automatically withdraw, however the Ball Valve WILL NOT automatically close. The Shear Valve can be manually actuated to establish containment integrity.

C. The TIP detector WILL NOT automatically withdraw and the Ball Valve WILL NOT automatically close. The Ball Valve can be manually actuated to establish containment integrity.

D. The TIP detector WILL NOT automatically withdraw and the Ball Valve WILL NOT automatically close. The Shear Valve can be manually actuated to establish containment integrity.

Answer: D ILO 2018 NRC Exam Page: 172 of 282 Question 58

Answer Explanation:

Failure of the PCIS to detect and/or initiate an isolation signal to Group 15 will result in the TIP detector remaining at its in-core location and failure of the TIP Ball Valve to close. Containment Integrity can be restored by manually actuating the TIP Shear Valve assembly, and is procedurally driven by 23.606, Section 6.2, Step 6.2.3.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly assume that, since the Ball Valve is a Primary Containment Isolation Valve, that it is the only component in the system that will not operate due to failure of the PCIS. This assumption is plausible and could lead the candidate to determine that the detector drive assembly will retract the detector and the Ball Valve will remain open, requiring manual action to close the Ball Valve and restore Containment Integrity.

B. The candidate could incorrectly assume that, since the Ball Valve is a Primary Containment Isolation Valve, that it is the only component in the system that will not operate due to failure of the PCIS. This assumption is plausible and could lead the candidate to determine that the detector drive assembly will retract the detector and the Ball Valve will remain open, requiring manual action to actuate the Shear Valve and restore Containment Integrity.

C. The candidate could correctly determine that the failure of the PCIS impacts both the detector retract circuit and the Ball Valve automatic isolation logic. However, the candidate could incorrectly determine that the Ball Valve could still be manually closed to restore Containment Integrity, which is plausible since the Ball Valve normally provides Containment Integrity when the TIP detector retracts into the shield chamber.

Reference Information:

SOP 23.606, Traversing In-Core Probe (TIP) System Student Text ST-OP-315-0025 NUREG 1123 KA Catalog Rev. 2 215001 K6. Knowledge of the effect that a loss or malfunction of the following will have on the TRAVERSING IN-CORE PROBE:

215001 K6.04 3.1/3.4 Primary containment isolation system: Mark-I&II(Not-BWR1) 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Traversing In-Core Probe (C5102)

Cognitive Enabler Discuss the Traversing Incore Probe interrelationships with other systems.

Traversing In-Core Probe (C5102)

Cognitive Enabler Discuss the Traversing Incore Probe interrelationships with other systems.

ILO 2018 NRC Exam Page: 173 of 282 Question 58

59 K/A Importance: 3.3/3.1 Points: 1.00 R59 Difficulty: 3.00 Level of Knowledge: High Source: NEW 73426 The plant at 100% power when the following is observed by the P603 Operator:

What is the impact to DCS logic, and what would the operator accomplish by taking the Reactor Level Select switch to B LEVEL?

A. Reactor Feed Pump speed will increase. The action could prevent a Level 8 trip if taken before the trip occurs.

B. Water level control will transfer to the B level instrument. The action would ensure the panel switch is aligned with DCS.

C. DCS will transfer to 1-element level control. The action would cause DCS to automatically transfer back to 3-element control.

D. Water level control will transfer to the median signal from the three remaining signals. The action would force DCS to use B as the lead level instrument.

Answer: D ILO 2018 NRC Exam Page: 174 of 282 Question 59

Answer Explanation:

Four narrow-range RPV water level signals, A, B, C, and D, are supplied to DCS logic. Operation of Reactor Level Select switch selects the lead RPV water level signal A or B, to control RPV water level and input to C32-R614, RPV Narrow Range Level Recorder. If the lead selected RPV water level signal, A or B fails, DCS logic will automatically transfer RPV water level control and the input to C32-R614, to the median signal from the three remaining RPV water level signals, preventing an RPV water level transient.

When a single RPV water level signal fails, the operator can restore control to a lead-selected RPV level instrument by changing the position of the selector switch. The Reactor Level Select switch position may be changed without causing an RPV water level transient.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This answer is plausible because the candidate could assume that the low RPV water level sensed by the lead-selected level signal would cause a level error (indicated level lower than level setpoint of 197) that would result in DCS logic sending a signal to the RFP speed control circuit to raise RFP speed to correct the level error. This is an incorrect assumption because of the design of DCS logic described above. If DCS logic behaved as described in the first part of this distractor, then selecting the B level to terminate the transient would be the correct action. Note that this action IS required if three level instruments were to fail and the remaining level signal can be selected as the lead signal.

B. This answer is plausible because there are only two level signals capable of being selected by DCS (A or B). The candidate could incorrectly assume that a failure of the lead-selected level signal would result in DCS logic transferring control to the non-selected level signal. However, this assumption is incorrect because of the design of DCS logic described above. If DCS responded as stated in this distractor, then placing the switch to B would be correct so that the switch position aligned properly with the signal being used by DCS for level control.

C. The answer is plausible because there are numerous failures (many of which have been on previous NRC exam questions) that would cause DCS to automatically transfer to 1-element (or single-element) control. The candidate could incorrectly recall this failure as being one of those. However, this is incorrect because of the design of DCS logic described above. IF DCS responded as stated in this distractor, then the correct action would be to change the lead-selected RPV water level signal to restore the preferred 3-element control.

Reference Information:

23.107, Reactor Feed and Condensate System, section 1.1 Feedwater System Description.

Student Text ST-OP-315-0046, Page 15, Instrumentation and Controls section for RPV Level NUREG 1123 KA Catalog Rev. 2 216000 A4. Ability to manually operate and/or monitor in thecontrol room:

216000 A4.02 3.3/3.1 Channel select controls 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Reactor Pressure Vessel Instrumentation (B2100)

Cognitive Terminal Given various controls and indications for Reactor Pressure Vessel Instrumentation system operations, analyze that indication to determine proper plant response in accordance with approved plant procedures.

ILO 2018 NRC Exam Page: 175 of 282 Question 59

60 K/A Importance: 3.1/3.1 Points: 1.00 R60 Difficulty: 4.00 Level of Knowledge: High Source: NEW 76069 The plant has been Emergency Depressurized due to a LOCA / Loss of injection.

• RHR Pumps A and B are injecting to the RPV at 12,500 gpm each through Div 2 RHR.

• The E1150-F010, RHR Crosstie Header Valve, is OPEN.

• Division 1 and 2 RHR Pump Discharge Pressures are as indicated.

Which of the following indicates the change in RHR system pressure that will occur due to initiating Drywell Sprays with Division 1 RHR?

A.

ILO 2018 NRC Exam Page: 176 of 282 Question 60

B.

C.

D.

Answer: B ILO 2018 NRC Exam Page: 177 of 282 Question 60

Answer Explanation:

Note: The information used to develop the graphics in this question was obtained from 29.ESP.01, Supplemental Information, Section 16.0, Pump Capacities Table, surveillance paperwork (such as 24.204.01 and 24.206.06, Div 1 and Div 2 LPCI and Suppression Pool Cooling/Spray Pump and Valve Operability Test) and the RHR system student text.

The RHR pumps develop 230 psig discharge pressure at 10,000 gpm flow. As pump flow increases, discharge pressure decreases. The starting conditions in the stem of the question were 12,500 gpm per pump, which results in a discharge pressure of ~205 psig.

With the E1150-F010 open, system flow (for any use) will be supported by both divisions of RHR. When an additional flow path is established by aligning Drywell Sprays with Division 1 RHR, because the E1150-F010 is open, both Division 1 and 2 pump discharge pressures will lower as pump pressure is converted to increased flow. The conditions shown in Answer B result from both pumps carrying ~17,500 gpm of flow at the shown discharge pressure of ~165 psig.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor shows Pump A and B discharge pressures staying the same as in the initial conditions.

This distractor is plausible because the candidate could fail to recall the relationship between increasing system flow, by opening another flow path through the Drywell Spray line, and pump discharge pressure. The candidate could incorrectly conclude that pump discharge pressure is a fixed value and will remain at the value shown in the stem (~205 psig) when Drywell Sprays is initiated.

C. This distractor shows Pump A discharge pressure increasing to ~235 psig. This indication was developed by closing the E1150-F010, RHR Crosstie Valve, and then placing Division 1 RHR in Drywell Sprays, by itself. This would result in 10,000 gpm flow at 235 psig. This distractor is plausible because the SRO has the option of closing the E1150-F010 to split out divisions of RHR to use one division for containment cooling while leaving the other division in the LPCI mode. If the candidate assumes that the E1150-F010 was closed, contrary to the information stated in the stem of the question, or if the candidate assumes that the E1150-F010 is always closed when spraying the drywell, which is not the case, then the candidate may recognize this discharge pressure shown as being consistent with one pump at 10,000 gpm (nominal drywell spray flow).

D. This distractor shows Pump B discharge pressure staying the same as in the initial conditions. This distractor is plausible because the candidate could fail to recall the relationship between increasing system flow, by opening another flow path through the Drywell Spray line, and pump discharge pressure. This distractor is also plausible because the candidate could assume that, since Division 1 RHR is being placed in Drywell Sprays, then division 2, or Pump B, would remain unaffected. Or, the candidate could have missed the fact that the E1150-F010 is open and incorrectly assume that Div 2 RHR would remain unaffected by Division 1 being placed in Drywell Sprays.

Reference Information:

E11-00, RHR Design Basis Document (DBD).

29.ESP.01, Supplemental Information, Section 16.0, Pump Capacities Table.

24.204.01, Div 1 LPCI and Suppression Pool Cooling/Spray Pump and Valve Operability Test.

24.204.06, Div 2 LPCI and Suppression Pool Cooling/Spray Pump and Valve Operability Test.

ILO 2018 NRC Exam Page: 178 of 282 Question 60

NUREG 1123 KA Catalog Rev. 2 226001 A1. Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI: CONTAINMENT SPRAY SYSTEM MODE controls including:

226001 A1.07 3.1/3.1 System pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Residual Heat Removal (E1100)

Cognitive Terminal Given various controls and indications for RHR system operations, analyze that indication to determine proper plant response in accordance with approved plant procedures.

ILO 2018 NRC Exam Page: 179 of 282 Question 60

61 K/A Importance: Points: 1.00 R61 Difficulty: 0.00 Level of Knowledge: Low Source: BANK 75927 Select the response that correctly completes the following statement.

The refueling interlocks ensure that __(1)__ does not occur during fuel handling operations by preventing __(2)__ with the Mode Switch in the Refuel position.

A. (1) inadvertent criticality; (2) control rod withdrawal whenever loaded fuel equipment is over the core B. (1) inadvertent criticality; (2) control rod withdrawal whenever fuel loading equipment is energized C. (1) excessive iodine gas release; (2) refueling hoist movement when RPV Water Level is <20' 6" above the top of the RPV flange D. (1) excessive iodine gas release; (2) withdrawl of any control rods when RPV Water Level is <20' 6" above the top of the RPV flange Answer: A ILO 2018 NRC Exam Page: 180 of 282 Question 61

Answer Explanation:

The refueling interlocks are explicitly assumed in the UFSAR SAFETY ANALYSES for the control rod removal error during refueling and the fuel assembly insertion error during refueling. These analyses evaluate the consequences of control rod withdrawal during refueling and also fuel assembly insertion with a control rod withdrawn. A prompt reactivity excursion during refueling could potentially result in fuel failure with subsequent release of radioactive material to the environment. Criticality and, therefore, subsequent prompt reactivity excursions are prevented during the insertion of fuel, provided all control rods are fully inserted during the fuel insertion. The refueling interlocks accomplish this by preventing loading of fuel into the core with any control rod withdrawn or by preventing withdrawal of a rod from the core during fuel loading. The refueling platform location switches activate at a point outside of the reactor core such that, considering switch hysteresis and maximum platform momentum toward the core at the time of power loss with a fuel assembly loaded and a control rod withdrawn, the fuel is not over the core.

Therefore, the candidate should recognize that the refuleing interlocks (1) prevent inadvertent criticality by (2) preventing control rod withdrawal whenever loaded fuel equipment is over the core.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could correctly recall that the refueling interlocks (1) prevent inadvertent criticality.

However, the candidate could incorrectly recall HOW this interlock is imposed by the logic and (2) conclude that it monitors status of power to the fuel loading equipment.

C. TS LCO 3.9.6, RPV Water Level during Refueling Operations requires that, during the movement of irradiated fuel assemblies within the RPV cavity, or handling of new fuel or control rods within the RPV while irradiated fuel assemblies are seated in the RPV, a minimum water level of 20 ft 6 inches above the top of the RPV flange be maintained to ensure an adequate water level in the reactor vessel cavity and spent fuel pool, which is necessary to retain sufficient iodine fission product activity assumed to be released in the water in the event of a fuel handling accident. The candidate could incorrectly determine that this (1) TS requirement for RPV Water Level is also interlocked with (2) refueling hoist equipment when the Mode Switch is in the Refuel position, which is incorrect.

D. TS LCO 3.9.6, RPV Water Level during Refueling Operations requires that, during the movement of irradiated fuel assemblies within the RPV cavity, or handling of new fuel or control rods within the RPV while irradiated fuel assemblies are seated in the RPV, a minimum water level of 20 ft 6 inches above the top of the RPV flange be maintained to ensure an adequate water level in the reactor vessel cavity and spent fuel pool, which is necessary to retain sufficient iodine fission product activity assumed to be released in the water in the event of a fuel handling accident. The candidate could incorrectly determine that this (1) TS requirement for RPV Water Level is also interlocked with (2) control rod movement when the Mode Switch is in the Refuel position, which is incorrect.

Reference Information:

TS Bases 3.9.1, 3,9.6 (for Distractors C and D).

NUREG 1123 KA Catalog Rev. 2 234000 Fuel Handling G2.1.28 4.1/4.1 Knowledge of the purpose and function of major system components and controls 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK Low RO ILO 2018 NRC Exam Page: 181 of 282 Question 61

Associated objective(s):

Refueling (F1500)

Cognitive Enabler Discuss the function and purpose of Refueling System components, including their importance to nuclear safety.

Refueling (F1500)

Cognitive Enabler List the interlocks associated with Refueling System components.

Refueling (F1500)

Cognitive Enabler Discuss the function and purpose of Refueling System components, including their importance to nuclear safety.

ILO 2018 NRC Exam Page: 182 of 282 Question 61

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 183 of 282 Question 62

62 K/A Importance: 2.7/2.7 Points: 1.00 R62 Difficulty: 3.00 Level of Knowledge: Low Source: NEW 73526 A ground on the BOP DC electrical distribution system occurs and causes the following:

• 2PC3-17 Position 1 to TRIP

• 4D49, AVR 125V DC Supply Trouble to ALARM What is the effect of this DC loss on the Main Turbine Generator?

A. DC voltage supplied to the Main Generator field will be lost.

B. The ability to regulate the Main Generator field will be disabled.

C. If a Main Turbine trip occurred, the generator field breaker would fail to trip.

D. The Main Turbine would coast down without an adequate lube oil supply if the AC Lube Oil pumps were lost.

Answer: B ILO 2018 NRC Exam Page: 184 of 282 Question 62

Answer Explanation:

2PC3-17, Position 1 is the DC electrical distribution power supply to the Main Turbine Generator Excitation Control Cubicle. Loss of this source results in a loss of power to circuits Q80 and Q40. Circuit Q40 supplies power to the binary inputs for the Main Turbine Generator Voltage Regulator (AVR) which regulates the amount of DC voltage applied to the generator rotor for supplying and regulating the DC field and, therefore, DC output voltage or reactive load. Loss of these controls renders the voltage regulator inoperable so that automatic and manual voltage regulation is lost.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor is plausible if the candidate thought that this DC circuit supplied voltage to the generator field. This is incorrect since generator field DC voltage comes from the output of the Power Convertor Thyristor Bank. This distractor is also plausible if the candidate confused this power supply loss with a loss of all of 2PC3-17, which would cause a loss of the generator field and a Main Generator trip.

C. This distractor is plausible if the candidate assumed that this circuit also powered the generator field breaker control circuit. This is incorrect because the Field Breaker DC Control circuit is powered from 2PC3-16.

D. This distractor is plausible if the candidate assumed that this circuit also powered the Main Turbine Emergency Oil Pump or its control circuit. This is incorrect because the Emergency Oil Pump is powered from 2PC-1, Pos 4A, which is the same power supply to the pump's DC control circuit.

Reference Information:

ARP 4D49, AVR 125V DC Supply Trouble Student Text ST-OP-315-0055 NUREG 1123 KA Catalog Rev. 2 245000 Main Turbine Generator and Auxiliary System 245000 K1. Knowledge of the physical connections and/or causeeffect relationships between MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS and the following:

245000 K1.09 2.7/2.7 D. C . electrical distribution 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (4) Secondary coolant and auxiliary systems that affect the facility.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Main Generator and Excitation (N3031 & N3034)

Cognitive Enabler Describe the normal and alternate power supplies to Main Generator and Excitation System components.

Main Generator and Excitation (N3031 & N3034)

Cognitive Enabler Describe the normal and alternate power supplies to Main Generator and Excitation System components.

ILO 2018 NRC Exam Page: 185 of 282 Question 62

63 K/A Importance: 3.3/3.3 Points: 1.00 R63 Difficulty: 3.00 Level of Knowledge: High Source: BANK 73546 The plant was operating at rated conditions when a STATION BLACKOUT occurred.

Shortly after the transient, the CRLNO reported that the Bus 1-2 red power available light was lit.

Which of the following describes the AVAILABILITY of power, and/or the QUICKEST method to restore power, to the SBFW system to allow injection to the RPV?

A. Power CAN NOT be restored to the SBFW pumps.

B. Power CAN be restored to the SBFW pumps by closing the A6 breaker for Transformer 64.

C. Power CAN be restored to the SBFW pumps by starting CTG 11 Unit 1 and restoring power to Transformer 64.

D. Power is CURRENTLY AVAILABLE to the SBFW pumps, no further actions are necessary to restore power for injection.

Answer: B ILO 2018 NRC Exam Page: 186 of 282 Question 63

Answer Explanation:

The SBFW pumps are motor driven feedwater pumps that serve as a High-Pressure injection source of water to the RPV. The pumps are powered from 4160V Busses 64V (East, or A) and 65W (West, or B).

Busses 64V and 65W can be cross-tied such that both pumps can be powered from either the Division 1 (120 kV) or Division 2 (345kV) offsite circuits.

The conditions given in the stem of the question indicate that at least one 120kV offsite circuit is available and currently powering Bus 1-2, as indicated by the red Bus 1-2 red power available light being lit. With power available at Bus 1-2, the operator can readily power transformer 64 by closing in the A6 breaker.

Powering transformer 64 restores power to bus 64V, which restores power to the SBFW pumps.

Note: At Fermi 2, the Station Blackout procedure has a unique situation that differentiates it from other stations. At Fermi 2, it is possible to be in the Station Blackout procedure but still have power available at Bus 1-2. If this is the case, power can be restored to the SBFW pumps as specified in the correct answer.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor is plausible if the candidate fails to correctly recall the flow of power from Division 1 (120kV) offsite power, through the distribution system, to transformer 64 and eventually busses 64V and 65W. The information in the stem of the question could lead the candidate to believe that power is not available, and therefore cannot be restored, to the SBFW pump motors.

C. This distractor is plausible if the candidate fails to correctly recall the flow of power from Division 1 (120kV) offsite power, through the distribution system, to transformer 64 and eventually busses 64V and 65W. Starting CTG 11-1 is an alternative to closing in the A6 position. However, starting CTG 11-1 is not necessary because Bus 1-2 is already energized, as stated in the stem of the question.

D. This distractor is plausible if the candidate fails to correctly recall the flow of power from Division 1 (120kV) offsite power, through the distribution system, to transformer 64 and eventually busses 64V and 65W. The information in the stem of the question could lead the candidate to believe that, with Bus 1-2 energized, power is already available to the SBFW pumps. This is not true, however, without first closing the A6 breaker.

Reference Information:

SD-2500-01, One Line Diagram Plant 4160V & 480V System Service Student Text ST-OP-315-0018, SBFW System NUREG 1123 KA Catalog Rev. 2 259001 Reactor Feedwater System 259001 K2. Knowledge of electrical power supplies to the following:

259001 K2.01 Reactor feedwater pump(s): Motor-Driven-Only 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK High RO Associated objective(s):

Standby Feedwater (N2103)

Cognitive Enabler Describe the normal and alternate power supplies to Standby Feedwater System components.

Standby Feedwater (N2103)

Cognitive Enabler Describe the normal and alternate power supplies to Standby Feedwater System components.

ILO 2018 NRC Exam Page: 187 of 282 Question 63

64 K/A Importance: 2.9/3.3 Points: 1.00 R64 Difficulty: 4.00 Level of Knowledge: High Source: NEW 73806 The plant is operating at 100% power when the following sequence of events takes place:

• 0700 Indications of a fire in the Cable Spreading Room are received in the Main Control Room. As a result, the following radiation monitors fail downscale at the times specified:

• 0730 K11-K609A, Div 1 Fuel Pool East Vent Exh Duct Rad Monitor.

• 0800 K11-K609C, Div 1 Fuel Pool West Vent Exh Duct Rad Monitor.

• 0810 K11-K609B, Div 2 Fuel Pool East Vent Exh Duct Rad Monitor.

• 0830 K11-K609D, Div 2 Fuel Pool West Vent Exh Duct Rad Monitor.

Which of the following identifies the EARLIEST TIME that CCHVAC will shift to the Recirculation Mode?

A. 0730 B. 0800 C. 0810 D. 0830 Answer: C ILO 2018 NRC Exam Page: 188 of 282 Question 64

Answer Explanation:

Per ARP 3D27, A downscale failure of channels A and B or C and D will cause CCHVAC to shift to the Recirculation Mode. The earliest time that this logic is satisfied is at 0810 when the A and B (and C, although it is not needed to satisfy the logic) instruments have failed.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly recall the logic for secondary containment isolation on a Fuel Pool radiation monitor failure. CCHVAC will also shift to Recirc if just ONE Fuel Pool rad monitor goes UPSCALE. If the candidate assumed that the downscale logic was the same, then this distractor could be chosen.

B. The candidate could incorrectly recall the logic for secondary containment isolation on a Fuel Pool radiation monitor failure. The candidate could assume that the logic was a downscale failure of channels A and C or B and D. Since both A and C instruments are Division 1 instruments, this line of thinking is plausible so this distractor could be chosen.

D. The candidate could incorrectly recall the logic for secondary containment isolation on a Fuel Pool radiation monitor failure. The candidate could assume that the logic required a downscale failure of ALL Fuel Pool radiation monitors to shift CCHVAC to Recirc. This line of thinking is plausible because the logic does require BOTH rad monitors in the same duct (East or West) to fail downscale before the logic actuates. The candidate could choose this distractor by carrying this line of thinking to the entire system (all 4 rad monitors) rather than just one duct.

Reference Information:

ST-OP-315-0050, Process Radiation Monitoring System Student Text.

ARP 3D27, FP Vent Exh Rad Monitor Downscale/INOP NUREG 1123 KA Catalog Rev. 2 272000 Radiation Monitoring System 272000 K3. Knowledge of the effect that a loss or malfunction of the RADIATION MONITORING System will have on following:

272000 K3.10 Control room ventilation: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Process Radiation Monitoring System (D1100)

Cognitive Enabler Discuss the Process Radiation Monitoring System interrelationships with other systems.

ILO 2018 NRC Exam Page: 189 of 282 Question 64

65 K/A Importance: 3.5/3.8 Points: 1.00 R65 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 74986 The plant is operating at 100% power with the following conditions:

• You are the CRLNO.

• Movement of material is taking place from the Reactor Building into the Railroad Airlock.

• 8D24, Rail Airlock Inner Door Seal Pressure Low is IN ALARM.

• 17D43, Rail Airlock Outer Door Seal Pressure Low is CLEAR.

A contractor calls you on the radio and says that he is in the Airlock going from the Reactor Building Fifth Floor to the Auxiliary Building and CANNOT open the Auxiliary Building airlock door.

What is the cause for his not being able to exit into the Auxiliary Building?

A. Power to the door interlock is de-energized.

B. The door from the Reactor Building is not closed.

C. The seals for the Reactor Building Door are not pressurized.

D. The seals for the Auxiliary Building Door are not depressurized.

Answer: B ILO 2018 NRC Exam Page: 190 of 282 Question 65

Answer Explanation:

The correct answer describes the airlock and door arrangement that someone would use to access AB-5, from RB-5, as asked in the stem of the question. The airlock consists of two airtight doors separated by a vestibule. Electrical interlocks prevent both doors from being simultaneously open. Entry to and from Secondary Containment, through this airlock, is accomplished without uncontrolled release of Secondary Containment atmosphere to the environment.

The candidate should recognize this type of door/interlock arrangement and determine that the RB door may not be closed, thereby preventing operation of the AB door.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor is plausible if the candidate recalled that the door interlocks were in effect if power to them was lost, which is incorrect. If de-energized, the interlocks are defeated meaning both doors can be open at the same time, as can be seen by Precaution and Limitation 3.2 in 23.428, Secondary Containment Airlocks and Penetrations.

C. This distractor is plausible if the candidate confused the door in the stem of the question with the type of airlock used for equipment access to and/or from the Reactor Building via the Rail Airlock doors.

This door type is equipped with inflatable seals that prevent the free movement of air in/out of the Reactor Building. It is plausible that the candidate could confuse 8D24 being in alarm with the door in the stem of the question and determine that the reason the outer (Auxiliary Building) door cannot be opened is because the inner (Reactor Building) door seal is depressurized.

D. This distractor is plausible if the candidate confused the door in the stem of the question with the type of airlock used for equipment access to and/or from the Reactor Building via the Rail Airlock doors.

This door type is equipped with inflatable seals that prevent the free movement of air in/out of the Reactor Building. It is plausible that the candidate could confuse 17D43 being clear with the door in the stem of the question and determine that the reason the outer (Auxiliary Building) door cannot be opened is because the Auxiliary Building door seal is still pressurized.

Reference Information:

23.428, Secondary Containment Airlocks and Penetrations Student Text ST-OP-315-0016, Containment Systems NUREG 1123 KA Catalog Rev. 2 290001 Secondary Containment 290001 K4. Knowledge of SECONDARY CONTAINMENT design feature(s) and/or interlocks which provide for the following:

290001 K4.01 3.5/3.8 Personnel access without breaching secondary containment: Plant-System 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Containment Systems (T2200 & T2300)

Cognitive Terminal Given various controls and indications for Containment Systems operations, analyze that indication to determine proper plant response in accordance with approved plant procedures.

ILO 2018 NRC Exam Page: 191 of 282 Question 65

66 K/A Importance: 4.4/4.7 Points: 1.00 R66 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED 73829 The plant was operating at 100% power with Jet Pump Total Flow of 91.7 Mlbm/hr when reactor power lowered. The following flow indications were observed:

Based ONLY on the indications above, what event has occurred?

A. Jet Pump Failure on Loop A B. Jet Pump Failure on Loop B C. Uncontrolled Recirc Flow change of 5% on A Reactor Recirc Pump (lowering)

D. Uncontrolled Recirc Flow change of 5% on B Reactor Recirc Pump (rising)

Answer: A ILO 2018 NRC Exam Page: 192 of 282 Question 66

ILO 2018 NRC Exam Page: 193 of 282 Question 66 Answer Explanation:

20.138.02 Jet Pump Failure Symptoms:

• Unexplained change in indicated Core Flow

• Unexplained change in Recirc Loop Flow

• Unexplained decrease in Core D/P

• Jet Pump Percent Differential Pressure deviates excessively from the average of the remaining Jet Pump Percent Differential Pressures The increase in Recirc Loop Flow in the A Loop compared to B Loop (B31-R617 vs B31-R613) and the decrease in Jet Pump Flow in the A Loop vs the B loop (B21-R611A vs B21R611B) indicates that a break in a Jet Pump in the A Loop has occurred, which has caused an increase in Recirc Loop Flow (due to a reduction in back pressure on the RR A RR Pump) and a decrease in Jet Pump Flow (due to flow bypassing the jet pumps).

Distracter Explanation:

B. This distractor is plausible because the candidates could misread the Jet Pump Flow indications and apply the deviation in Jet Pump Flows to the B Loop, vice the A Loop, escpecially since, at Fermi 2, Jet Pumps 1 through 10 are in the A Loop and Jet Pumps 11 through 20 are in the B Loop. This answer is incorrect because the difference between driving (B31-R617) and driven (B21-R611) flow in the A loop indicates that a break has occurred in the A loop. B loop Jetpump flows as indicated by B21-R609B, B21-R609D and B21-R611B all indicate consistent flows and do not provide indication of a Jet Pump failure in the B Loop.

C. This distractor is plausible because Jet Pump Flow on the A loop has decreased and is about 10%

lower than the B loop (B21-R611A vs B21R611B), which could lead the candidate to conclude that A Loop Recirc Flow has lowered. This is incorrect, however, because the imbalance between Jet Pump Flows and Recirc Loop Flows as well as the deviation between Calibrated Jet Pump Flows (B21-R609A & C vs B21-R611A) indicate that a Jet Pump Failure has occurred, and not a Recirc Flow decrease.

D. This distractor is plausible because Jet Pump Flow on the B loop is about 10% higher than Jet Pump Flow on the A Loop (B21-R611A vs B21R611B), which could lead the candidate to conclude that B Loop Recirc Flow has increased. This is incorrect, however, because the stem of the question states that reactor power has lowered, which would not occur on a Recirc Flow increase. Also, the imbalance between Jet Pump Flows and Recirc Loop Flows in the A Loop, as well as the deviation between Calibrated Jet Pump Flows (B21-R609A & C vs B21-R611A) indicate that a Jet Pump Failure has occurred, and not a Recirc Flow increase.

Reference Information:

AOP 20.138.02 Jet Pump Failure (pg 5) Jet Pump Failure Symptoms ILO 2018 NRC Exam Page: 194 of 282 Question 66

Plant Procedures 23.138.02 NUREG 1123 KA Catalog Rev. 2 G2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage ILO 2015 Exam NRC Question Use (ILO 2018)

High MODIFIED RO Associated objective(s):

Cycle 15-3 Objectives Cognitive Terminal Given a copy of 20.138.02, discuss licensed operator actions for Jet Pump Failure ILO 2018 NRC Exam Page: 195 of 282 Question 66

67 K/A Importance: 4.6/4.3 Points: 1.00 R67 Difficulty: 4.00 Level of Knowledge: High Source: NEW 74987 NOTE: Use the SBFW System Mimic handout to answer this question.

The plant is at 100% power in MODE 1. The East SBFW Pump was just shut down following the performance of 24.107.03, SBFW Pump and Valve Operability and Lineup Verification Test.

You have been tasked with verifying that the system is in the standby lineup.

Which of the responses below accurately completes the following statement to reflect the standby availability of the SBFW system?

If the CMC Switches for the SBFW Pumps are taken to RUN, __(1)__ will start and __(2)__ inject into the RPV.

A. (1) ONLY the West (2) WILL B. (1) BOTH (2) WILL C. (1) BOTH (2) NOT D. (1) ONLY the East (2) NOT Answer: D ILO 2018 NRC Exam Page: 196 of 282 Question 67

Answer Explanation:

Following completion of the flow verification test, the SBFW Auxiliary Lube Oil pump is left running for 5 minutes. During this time, however, the Mode Selection switch should be in NORM. With the switch in TEST, as shown in the picture of the system mimic, starting logic for the SBFW pumps and injection valve is impacted. When the Mode Selection switch is in TEST, the N2103-F001, SBFW Disch to RPV Iso Valve, will NOT open. Normally, this valve would open automatically when the first pump is started.

Additionally, when the Mode Selection switch is in TEST, the SBFW pumps are prevented from running unless an oil pressure permissive is met by running the pump's Auxiliary Lube Oil Pump.

With this information, the candidate should recognize that ONLY the East SBFW Pump will run when it's CMC is taken to RUN. The candidate should also recognize that the N2103-F001 will not open preventing injection into the RPV when in TEST.

Note: The Mode Selection switch being in TEST does NOT generate any alarms in the Main Control Room to inform the operator of the switch being out of position, so it is important for the candidate to recognize this switch being out of its normal standby position and to understand the switch's impact on system logic.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly apply precaution 3.2 of the SBFW System SOP, which states that "When the SBFW System is aligned in the Test mode, only one pump is to be tested at a time". The candidate could assume that this precaution describes logic that is enforced by the Mode Selection switch being in TEST that would only allow one pump to start when in TEST, which is not correct. The candidate could also fail to recall the impact of the Mode Selection switch on the logic for the N2103-F001 valve or fail to recognize that the switch is in TEST.

B. The candidate could fail to recognize that the Mode Selection switch is in TEST, which would lead the candidate to believe that both pumps will start and the N2103-F001 will open on pump start. Also, the candidate could recognize the switch being in TEST but fail to recall the impact of this on the pump start logic and on the logic for the N2103-F001.

C. The candidate could recognize that the Mode Selection Switch is in the TEST position, but fail to recall the impact of this on pump start logic and conclude that both pumps will start when their CMCs are taken to RUN, which is not correct. The candidate could correctly recall the impact of the switch being in TEST on the logic for the N2103-F001.

Reference Information:

23.107.01, SBFW System 24.107.03, SBFW Pump and Valve Operability and Lineup Verification Test ST-OP-315-0018, Standby Feedwater System Student Text, Table 4 SBFW Control Functions and Interlocks NUREG 1123 KA Catalog Rev. 2 G2.1.31 Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

High NEW RO Associated objective(s):

Standby Feedwater (N2103)

Cognitive Terminal Given various controls and indications for SBFW operations, analyze that indication to determine proper plant response in accordance with approved plant procedures.

ILO 2018 NRC Exam Page: 197 of 282 Question 67

68 K/A Importance: 4.3/4.3 Points: 1.00 R68 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 73847 While operating at power, a reactor scram occurs.

Following the scram, alarm 2D28, HPCI System Actuated, is received.

With the RPV still pressurized to 1000 psig, which of the following RPV level indications should you use to validate that the HPCI system is responding to an actual plant condition versus a spurious actuation?

A. B21-R605, RPV Flood Up Level Indicator, reading 160 inches.

B. C32-R606A(B), Reactor Level A(B) Indicator(s), reading 160 inches.

C. B21-R604A(B), Div 1(2) Reactor Water Level Indicator(s), reading 100 inches.

D. B21-R610(615), Div 1(2) RPV Core Level Recorder(s), reading 100 inches.

Answer: C ILO 2018 NRC Exam Page: 198 of 282 Question 68

Answer Explanation:

The HPCI system is actuated on a low RPV water level condition when RPV level drops less than 110.8" as sensed by the Wide Range RPV Level instruments. Per ARP 2D28, HPCI System Actuated, this is verified by observing the B21-R604A(B) Div 1(2) Reactor Water Level Indicators showing less than 110.8".

These indicators are driven by Wide Range level instruments. The candidate should recognize that the indicators and value listed in the correct answer are driven by Wide Range instruments and are therefore indicators than can be used to validate proper response of the instruments that input into HPCI initiation logic.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The B21-R605, Flood Up Level indicator is low off scale and the candidate could interpret this as being a valid indication to use to verify HPCI system response. However, the lower end of the indicating range for this instrument is 160", so using it to validate an actuation at 110.8" is not acceptable. Additionally, this instrument is not related to the Wide Range level instruments that input into HPCI initiation logic and is not listed in ARP 2D28.

B. The C32-R606A(B), Reactor Level A(B) Indicators are low off scale and the candidate could interpret this as being a valid indication to use to verify HPCI system response. However, the lower end of the indicating range for these instruments is 160", so using them to validate an actuation at 110.8" is not acceptable. Additionally, these instruments are not related to the Wide Range level instruments that input into HPCI initiation logic and they are not listed in ARP 2D28.

D. The B21-R610(615), Div 1(2) RPV Core Level Recorder is reading below the initiation setpoint for the HPCI logic. However, the Core Level instruments are calibrated for a usable range of -150 to +50 inches. Although the recorder will display information above +50", it should not be used for validating HPCI system response. Additionally, these instruments are not related to the Wide Range level instruments that input into HPCI initiation logic and are not listed in ARP 2D28.

Reference Information:

ARP 2D28, HPCI System Actuated.

ST-OP-315-0021, RPV Instrumentation Student Text, Page 17 Monitoring Instrumentation.

NUREG 1123 KA Catalog Rev. 2 G2.1.45 4.3/4.3 Ability to identify and interpret diverse indications to validate the response of another indication 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Reactor Pressure Vessel Instrumentation (B2100)

Cognitive Enabler Discuss effective monitoring of Reactor Pressure Vessel Instrumentation using local, remote, computer displays and alarms.

ILO 2018 NRC Exam Page: 199 of 282 Question 68

69 K/A Importance: 3.7/4.1 Points: 1.00 R69V2 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 75106 A Licensed Nuclear Operator (LNO) is attempting to perform an Operations department Technical Specification surveillance procedure. He has encountered a non-conditional step that cannot be performed under the current plant conditions and the LNO believes that all the requirements of MGA03, Procedure Use and Adherence, are met to use Not Applicable (N/A) for this step.

Who can approve marking the step N/A?

A. Shift Manager B. Operations Engineer C. Field Support Supervisor D. Control Room Supervisor Answer: A ILO 2018 NRC Exam Page: 200 of 282 Question 69

Answer Explanation:

MGA03, step 5.1.10.3.c states that the Shift Manager must approve the use of N/A for all non-conditional steps for all Operations procedures and Technical Specification and Technical Requirements Manual Surveillances.

Therefore, the candidate must determine that Shift Manager approval is required in this instance since the individiual is an LNO who is attempting to perform an Operations department Tech Spec surveillance.

Distractor Explanation:

B. Is plausible if the candidate believes that a higher level of approval, beyond the Shift Manager, is required to waive non-conditional steps in Tech Spec related surveillances. This is plausible because other Operations Department administrative requirements, such as the approval required to not perform system lineups in 22.000.01, Plant Startup Master Checklist, does require the Operations Engineer to N/A steps.

C. Is plausible because the use of N/A in non-Operations department procedures can be approved by the individual's immediate supervisor which, for LNOs assigned to the Research Tagging Center, is the Field Support Supervisor.

D. Is plausible because the use of N/A in non-Operations department procedures can be approved by the individual's immediate supervisor which, for LNOs assigned to the Main Control Room, is the Control Room Supervisor.

Reference Information:

MAG03, Procedure Use and Adherence (5.1.10) 22.000.01, Plant Startup Master Checklist NUREG 1123 KA Catalog Rev. 2 G2.2.12 Knowledge of surveillance procedures 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Administrative Conduct Manuals - Licensed Operators/STA Performance Terminal Describe when partial revisions may be made to procedures and any requirements that must be met.

Administrative Conduct Manuals - Licensed Operators/STA Cognitive Terminal Discuss the administrative requirements and responsibilities of the Licensed Operator role Administrative Conduct Manuals - Licensed Operators/STA Performance Terminal Describe the differences between Continuous Use and Reference Use procedures.

ILO 2018 NRC Exam Page: 201 of 282 Question 69

70 K/A Importance: 3.9/4.3 Points: 1.00 R70 Difficulty: 2.50 Level of Knowledge: Low Source: NEW 75866 You are reviewing an electrical print. The review takes you to the relay tabulation shown below:

What is physically located between Terminals 5-6 and what is the state of the component located between Terminals 1-7 if component 5-6 is ENERGIZED?

A. (1) Relay coil; (2) Open B. (1) Relay contact; (2) Closed C. (1) Relay coil; (2) Closed D. (1) Relay contact; (2) Open Answer: C ILO 2018 NRC Exam Page: 202 of 282 Question 70

Answer Explanation:

From the I&C Print Reading student text: "The most common symbol in a control schematic is a relay. Its symbol is shown in Figure 3 [not provided, but the symbol is shown at the top left of the relay tabulation between terminals 5-6] with some representative contacts. Relays and their contacts are shown in their shelf state (de-energized). When the relay is energized, the contacts switch to the opposite state."

The candidate should determine that the K14A component located between Terminals 5-6 is the relay's coil. The relays shown in the tabulation given in the stem of the question are all the GE HMA relay type.

When this type of relay, or specifically relay K14A, is in its shelf (de-energized) state, the contacts are as shown in the tabulation. The contacts across relay terminals 1-7 and 2-8 are open and the contacts across terminals 7-3 and 8-4 are closed. If the relay were to become energized, as stated in the stem of the question, the contacts would change state thus causing the contacts across relay terminals 1-7 to close.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could correctly determine that the relay's coil is between Terminals 5-6. The candidate could then incorrectly determine that relay on this relay tabulation are shown in the energized state and conclude that contact 1-7 would be open when the relay is energized. This is incorrect because relays are shown in their de-energized state in the relay tabulation.

B. The candidate could incorrectly determine that a relay contct is shown between Terminals 5-6 instead of the relay coil.

D. The candidate could incorrectly determine that a relay contct is shown between Terminals 5-6 instead of the relay coil. The candidate could then incorrectly determine that relay on this relay tabulation are shown in the energized state and conclude that contact 1-7 would be open when the relay is energized. This is incorrect because relays are shown in their de-energized state in the relay tabulation.

Reference Information:

I-2215-02, Grid F-8, GE Type HMA Relay Tabulation ST-GN-175-0003-001, Print Reading Operations - I&C.

Objective Information:

ST-GN-175-0003-001, Print Reading Operations - I&C Objective 0008 Describe the standard method by which relays, contacts, and switches are electrically pictured on schematic drawings.

Objective 0009 Use a relay tabulation to determine the function of each contact on a given relay.

NUREG 1123 KA Catalog Rev. 2 G2.2.15 3.9/4.3 Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line-ups, tag-outs, etc.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

ILO 2018 NRC Exam Page: 203 of 282 Question 70

71 K/A Importance: 3.9/4.5 Points: 1.00 R71 Difficulty: 2.00 Level of Knowledge: Low Source: BANK 73887 The Plant is in Mode 5 with movement of recently irradiated fuel in progress on the Refueling Floor. The following alarms are received:

• 8D46, DIV I REACTOR BLDG PRESSURE HIGH/LOW

• 17D46, DIV II REACTOR BLDG PRESSURE HIGH/LOW Reactor Building (RB) Pressure is - 0.120 (minus 0.120) inches WC.

What Technical Specification action MUST be taken?

A. Suspend the fuel movement immediately.

B. Verify RBHVAC supply and exhaust fans operating within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

C. Verify at least one door in each RB access is closed within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. Start both divisions of Standby Gas Treatment System immediately.

Answer: A ILO 2018 NRC Exam Page: 204 of 282 Question 71

Answer Explanation:

The stem of the question shows RB pressure high (not enough vacuum) and outside of the allowable range for LCO 3.6.4.1, which requires 0.125 inch of vacuum water gauge.

The actions taken are verbatim from Condition D of the LCO and the Immediate Action is required RO level of knowledge.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could recognize the need to verify proper operation of the RBHVAC system and lose site of the Immediate TS action. The action specified is also an action required by the ARP, however it is not a Tech Spec required action. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time limit given is consistent with the knowledge level required of RO candidates.

C. The candidate could recognize the need to verify that each one door in each RB access is closed, per the ARP, and lose site of the Immediate TS action. Although the action specified is an action required by the ARP, it is not a Tech Spec required action. The 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> time limit given is consistent with the knowledge level required of RO candidates.

D. The candidate could recognize starting SGTS as an action of the ARP and a viable response to high RB pressure and lose site of the Immediate TS action. Although the action specified is an action required by the ARP, it would only be required if RBHVAC was not operating (more information needed than provided in the stem of the question) and is not a Tech Spec required action. The immediate time limit given is consistent with the knowledge level required of RO candidates.

Reference Information:

Technical Specifications 3.6.4.1 ARPs 8D46 (17D46) Div I (II) Reactor Building Pressure High/Low NUREG 1123 KA Catalog Rev. 2 G2.2.39 Knowledge of less than one hour technical specification action statements for systems.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

NRC Question Use (ILO 2018)

BANK Low RO Associated objective(s):

Containment Systems (T2200 & T2300)

Cognitive Enabler Identify Containment Systems related technical specifications, with emphasis on action statements requiring prompt actions (for example, one hour or less).

ILO 2018 NRC Exam Page: 205 of 282 Question 71

72 K/A Importance: 3.2/3.7 Points: 1.00 R72 Difficulty: 4.00 Level of Knowledge: High Source: NEW 73906 The plant is operating at 100% power with Division 2 CCHVAC in service. A DBA LOCA occurs which results in an offsite radiation release and the following:

• CCHVAC has automatically shifted to the Recirculation mode.

• The Division 2 CCHVAC Emergency Makeup Fan is NOT running.

• Neither the North nor the South Emergency Air intakes has received a Hi-Hi radiation trip signal.

• The Division 1 and Division 2 Emergency Air Intake Selector switches are in AUTO.

An operator sent to the Relay Room reports the following:

• D11-K836A, Div 1 S CCHVAC Emerg Air Inlet Rad Monitor is reading 48,000 cpm.

• D11-K837A, Div 1 N CCHVAC Emerg Air Inlet Rad Monitor is reading 33,000 cpm.

• D11-K836B, Div 2 S CCHVAC Emerg Air Inlet Rad Monitor is reading 45,000 cpm.

• D11-K837B, Div 2 N CCHVAC Emerg Air Inlet Rad Monitor is reading 31,000 cpm.

What action is required?

A. Start Division 1 and shut down Division 2 CCHVAC.

B. Shut down the Division 1 CCHVAC Emergency Makeup Fan.

C. Place the Emergency Air Intake Selector switches in NORTH.

D. Maintain the Emergency Air Intake Selector switches in AUTO.

Answer: C ILO 2018 NRC Exam Page: 206 of 282 Question 72

Answer Explanation:

When in the Recirc Mode with the Emergency Air Intake Selector switch in AUTO, the Control Center Emergency Air Inlet Radiation Monitor system logic will auto-isolate one air inlet path (N/S) which has a Hi-Hi trip following a five-minute sampling period (in effect for selecting the air inlet with the lowest radiation level). Dampers on the tripped side will be closed for one hour. Both sides will then reopen for five minutes while inlet air is resampled. Inlet selection based on trip/no-trip will then be repeated. If no Hi-Hi trip is present, both inlets will remain open since there is no need (or signal present) to isolate one side or the other.

However, per 23.413 Section 7.3 as well as ARPs 3D38 and 3D39, with CCHVAC shifted to the Recirc Mode, and with an offsite radiation release in progress, the operator should place the Emergency Air Intake Selector switches to either NORTH or SOUTH, selecting the intake with lowest indicated radiation level.

In this instance, the operator should select NORTH based on the radiation levels given in the stem of the question.

Note: This concept is significant because it is a time critical operator action that must be performed within 30 minutes (see M-5988).

Distractor Explanation:

A. The candidate could recognize that both Division 1 radiation monitors are reading higher than both Division 2 radiation monitors. The candidate could also recognize that only the Division 1 CCHVAC Emergency Makeup Fan is running. Either of these could lead the candidate to determine that CCHVAC needs realigning to Division 1 in service; however, plant conditions do not dictate shifting divisions of CCHVAC.

B. The candidate could incorrectly determine that the CCHVAC Emergency Makeup Fan in the non-operating division (Division 1) needs to be shut down. This is plausible because, when CCHVAC shifts to the Recirc Mode, both Emergency Makeup Fans start and, if both fans are running, the normal course of action is to shut down the fan in the non-operating division. However, per 23.413 Section 7.3, the operator should leave the one makeup fan (Div 1) running, even if it is in the non-operating division.

D. The candidate could incorrectly recall the actions that need to be taken if neither CCHVAC Emergency Makeup intake receives a Hi-Hi radiation signal when a release is in progress. This is plausible since neither CCHVAC Emergency Makeup intake radiation monitor is above the Hi-Hi trip setpoint, and in the case the logic would keep both inlets open, which could lead the candidate to believe that the intake selector switch needs to stay in AUTO. However, this is incorrect as per 23.413 and ARPs 3D38 and 3D39, which both require the operator to select the intake with the lower reading.

Reference Information:

23.413 CCHVAC System SOP, Section 7.3 ARPs 3D38(39) CCHVAC Emerg Air Div I(II) Rad Mon Trouble M-5988 Operator Time Critical Actions and Design Basis Sheet 1 NUREG 1123 KA Catalog Rev. 2 290003 Control Room HVAC G2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(12) Radiological safety principles and procedures.

NRC Question Use (ILO 2018)

High NEW RO ILO 2018 NRC Exam Page: 207 of 282 Question 72

Associated objective(s):

Control Center HVAC (T4102)

Cognitive Enabler List the interlocks associated with Control Center HVAC system components.

Reactor Operator Performance Enabler Perform proper system operations in accordance with System Operating Procedures (SOP).

ILO 2018 NRC Exam Page: 208 of 282 Question 72

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 209 of 282 Question 73

73 K/A Importance: 3.4/3.8 Points: 1.00 R73 Difficulty: 3.00 Level of Knowledge: Low Source: NEW 73926 The plant is in an ATWS condition and you are performing 29.ESP.03, Alternate Control Rod Insertion Methods. You are evaluating the following procedure sections:

• Section 3.0, Manual Control Rod Insertion.

• Section 4.0, Scram Reset and Manual Scram Re-initiation.

• Section 7.0, Vent Scram Air Header.

• Section 8.0, Vent CRD Over Piston Volumes.

Per 29.ESP.03, which of these sections, if performed, may result in the release of pressurized radioactive steam or water from the RPV?

A. Sections 3.0 and 7.0.

B. Sections 7.0 and 8.0.

C. Sections 3.0 and 4.0.

D. Sections 4.0 and 8.0.

Answer: D ILO 2018 NRC Exam Page: 210 of 282 Question 73

Answer Explanation:

29.ESP.03, Sections 4.0 and 8.0 contain cautions informing the operator that the performance of the steps associated with these sections may result in the release of pressurized radioactive steam or water from the RPV. No other sections of 29.ESP.03 contain this caution.

Distractor Explanation:

Note: 29.ESP.03, Sections 3.0, 7.0, and 8.0 all contain Cautions warning the operator of the potential for higher than normal radiation levels that may be experienced during their performance.

Distractors are incorrect and plausible because:

A. The candidate could recall the caution regarding higher than normal radiation levels from Sections 3.0 and 7.0 and assume that the reason these sections contain this caution is because they could result in the release of radioactive steam or water from the RPV, which is incorrect.

B. The candidate could recall the caution regarding higher than normal radiation levels from Section 7.0 and assume that the reason this section contains this caution is because it could result in the release of radioactive steam or water from the RPV, which is incorrect. Section 8.0 is one of the sections that has the potential for releasing radioactive steam or water from the RPV.

C. The candidate could recall the caution regarding higher than normal radiation levels from Section 3.0 and assume that the reason this section contains this caution is because it could result in the release of radioactive steam or water from the RPV, which is incorrect. Section 4.0 is one of the sections that has the potential for releasing radioactive steam or water from the RPV.

Reference Information:

29.ESP.03, Alternate Control Rod Insertion Methods.

ST-OP-802-3006-001, Emergency Support Procedures Student Text.

NUREG 1123 KA Catalog Rev. 2 G2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(12) Radiological safety principles and procedures.

NRC Question Use (ILO 2018)

Low NEW RO Associated objective(s):

Emergency Support Procedures Performance Terminal Describe the methods of manually inserting control rods during an Anticipated Transient Without Scram.

ILO 2018 NRC Exam Page: 211 of 282 Question 73

74 K/A Importance: 4.0/4.2 Points: 1.00 R74 Difficulty: 3.00 Level of Knowledge: Low Source: BANK 73927 Which of the following describes the basis behind operation of the discharge valve for a tripped Reactor Recirculation Pump as required by the subsequent actions of 20.138.01, Recirculation Pump Tripped AOP?

A. The valve is CLOSED to prevent thermal binding or pressure locking of the valve.

B. The valve is RE-OPENED to prevent thermal binding or pressure locking of the valve.

C. The valve is CLOSED to prevent the pump from fully seating on its thrust bearing following the trip.

D. The valve is RE-OPENED to prevent the pump from fully seating on its thrust bearing following the trip.

Answer: B ILO 2018 NRC Exam Page: 212 of 282 Question 74

Answer Explanation:

From the Reactor Recirculation Pump Trip AOP (20.138.01):

"The recirculation pump discharge valve for the pump that is tripped is closed to allow for flow coast down and to prevent the possibility of reverse flow through the pump. It also ensures that the pump will have seated on its thrust bearing following the trip In order to prevent any thermal binding or pressure locking, the discharge valve should be opened after approximately five minutes from the time that it was closed. This will ensure that enough time has elapsed with the valve closed to address the flow concerns and the seating on the thrust bearing, yet be re-opened prior to any thermal binding or pressure locking occurring."

The correct answer is the only one that contains the correct basis for the action being taken in that it states the valve is re-opened to prevent thermal binding or pressure locking.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The valve is closed per the AOP, however, the reason that it is closed is NOT to prevent thermal binding or pressure locking. The reason the valve is closed is to allow for flow coast down and ensure the pump will seat on its thrust bearing. The valve is re-opened to prevent thermal binding or pressure locking.

C. The valve is closed per the AOP, however, the reason that it is closed is NOT to prevent the pump from fully seating on its thrust bearing but, in fact, to ENSURE the pump does seat on its thrust bearing.

D. The valve is re-opened per the AOP, however, the reason that it is re-opened is NOT to prevent the pump from fully seating on its thrust bearing. The reason it is re-opened is to prevent the valve from thermally binding or pressure locking.

Reference Information:

20.138.01, Reactor Recirculation Pump Trip AOP 20.138.01 Bases, Reactor Recirculation Pump Trip AOP Bases NUREG 1123 KA Catalog Rev. 2 G2.4.11 Knowledge of abnormal condition procedures.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

BANK Low RO Associated objective(s):

Reactor Operator Exam Objectives Performance Enabler Explain bases for notes, cautions and overrides.

ILO 2018 NRC Exam Page: 213 of 282 Question 74

75 K/A Importance: 3.6/4.0 Points: 1.00 R75 Difficulty: 4.00 Level of Knowledge: High Source: BANK 74546 The plant is currently operating at 100% power in a black board condition when the CRLNO notices that the VAS display shows 3D19 as red, along with VAS Hardware System Trouble and MUX A & C FAILED.

GIVEN a copy of the Alarm Response Procedure for 3D19, which of the responses below correctly completes the following statement regarding the impact of these conditions on the VAS system?

This information indicates that the VAS system will incur ____(1)____of functionality and will result in _____(2)_____.

A. (1) loss (2) loss of redundancy for one half side of VAS I/O B. (1) no loss (2) loss of P601 through P805 window failures C. (1) loss (2) loss of P601 through P805 window failures D. (1) no loss (2) loss of redundancy for one half side of VAS I/O Answer: C ILO 2018 NRC Exam Page: 214 of 282 Question 75

Answer Explanation:

ARP 3D19 includes a flowchart on Page 6 that the candidate should refer to based on the information provided in the stem of the question.

Due to the VAS System Hardware Failed and the blackboard condition, the operator should answer N to the question "have a number of false alarm windows turned on?" This should lead the candidate to review the table to determine Data Acquisition Status. Given the failure of MUX A and C, the operator should review Action C, which is located on the next page (page 7). This page describes the impact of a loss of the redundant pair of MUX A&C.

A failure of this pair (or any redundant pair, which includes A&C or B&D) will result in a loss of VAS functionality and loss of alarm windows on the P601 through P805 panels.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly read the flowchart in the ARP and read Action B, which would lead the candidate to determine that the loss MUX A/C means that redundancy for one half side of VAS was lost, which is incorrect because the redundant pair has been lost (A and C MUX are redundant to one another).

B. The candidate could incorrectly read the flowchart in the ARP and determine that no loss of functionality occurred, which is incorrect as stated on Action C.

D. The candidate could incorrectly read the flowchart in the ARP and determine that no loss of functionality occurred, which is incorrect as stated on Action C.

Reference Information:

ARP 3D19, Annunciator System Trouble (provided as reference)

NUREG 1123 KA Catalog Rev. 2 G2.4.32 3.6/4.0 Knowledge of operator response to loss of all annunciators 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Question Use (ILO 2018)

BANK High NRC Early Review RO Associated objective(s):

Integrated Plant Computer Systems (IPCS) / Annunciator and Sequence Recorder System Cognitive Enabler Identify alarm response procedures associated with the Integrated Plant Computer System /

Annunciator and Sequence Recorder System.

Integrated Plant Computer Systems (IPCS) / Annunciator and Sequence Recorder System Cognitive Enabler Identify alarm response procedures associated with the Integrated Plant Computer System /

Annunciator and Sequence Recorder System.

ILO 2018 NRC Exam Page: 215 of 282 Question 75

76 K/A Importance: 3.3/4.0 Points: 1.00 S76V2 Difficulty: 3.00 Level of Knowledge: High Source: BANK 75046 You are the Control Room Supervisor (CRS) when the following conditions occur:

• The station experiences a Complete Loss of AC Power.

• The HPCI system is the ONLY available RPV injection source.

• Primary Containment conditions dictate that an Emergency Depressurization (ED) be performed.

Once the ED is commenced, (1) at what point, if any, will you direct the ED be terminated and (2) what is the basis for this action?

A. (1) ED is never terminated until the SRVs close on low steam pressure.

(2) To ensure the RPV is at the lowest possible energy state.

B. (1) When RPV pressure reaches ~200 psig.

(2) To ensure sufficient steam pressure to allow continued operation of HPCI for Adequate Core Cooling (ACC).

C. (1) When less than 5 SRVs are open and RPV pressure is greater than 50 psig above torus pressure.

(2) To ensure adequate steam flow through open SRVs to maintain Adequate Core Cooling (ACC).

D. (1) When RPV pressure reaches ~75 psig.

(2) To prevent recieving a low pressure isolation for HPCI.

Answer: B ILO 2018 NRC Exam Page: 216 of 282 Question 76

Answer Explanation:

ED is terminated when RPV pressure reaches ~200 psig to ensure sufficient steam pressure to allow continued operation of HPCI since the HPCI low pressure isolation occurs at 100 psig.

The candidate should determine the correct response, as directed by the EOPs for the given plant conditions, per ED-OR1, which states IF "It is anticipated that RPV depress will result in loss of inj required for ACC" THEN "Terminate RPV depress" AND "Control RPV press as low as practicable while maintaining injection required for ACC. Defeat isolations and exceed Off-Site release rate limits if necessary."

Note: The override that this question is based on was added as a result of the Extended Loss of AC Power (ELAP) event encountered at Fukishima.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOP actions and it is not related to immediate actions, and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (ED during a loss of AC power with only a steam driven feed source available) and then taking action based on the requirements of the EOPs.

Distractor Explanation:

Note: Distractor Structure is NOT 2 of 2, each is a separate statement. The use of the (1) and (2) is to help the examinee understand the answers and distractors.

A. "ED is never terminated until the SRVs close on low steam pressure, to ensure the RPV is at the lowest possible energy state" is a valid distractor since this is the normal end state of ED in conditions not involving a loss of AC power with only a steam driven feed source available.

C. "When less than 5 SRVs are open and RPV pressure is greater than 50 psig, To ensure adequate steam flow through open SRVs to maintain Adequate Core Cooling (ACC)." is a logical distractor because this another override in the ED leg, ED-OR2, which is an override that requires that ED be performed (or re-performed) if the RPV were to become re-pressurized. This answer is incorrect because it does not ensure ACC in conditions 0involving a loss of AC power with only a steam driven feed source available.

D. "ED is terminated when RPV pressure reaches ~75 psig To prevent receiving a low-pressure isolation for HPCI" is a logical distractor since 62 psig is the isolation setpoint for the RCIC system and the candidate could confuse the RCIC and HPCI low pressure isolations. Additionally, the candidate could assume that the HPCI low pressure is defeated while under these conditions, which it is not.

Even if it were isolated, per the HPCI Design Basis Document, the isolations are based on the minimum pressure for HPCI turbine operation, so going below 100 psig would result in a loss of the HPCI turbine and a loss of ACC.

Reference Information:

29.100.01 SH 3 ED-OR1 HPCI DBD E41-00 ILO 2018 NRC Exam Page: 217 of 282 Question 76

Plant Procedures 29.100.01 SH 3 NUREG 1123 KA Catalog Rev. 2 295003 AK1. Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER :

295003 AK1.06 Station blackout: Plant-Specific G2.4.18 Knowledge of the specific bases for EOPs 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

BANK High SRO Associated objective(s):

EOP Objectives Cognitive Terminal Given a copy of 29.100.01 sheet 3, discuss the operator's actions for Reactor Pressure Vessel Emergency Depressurization (ED)

ILO 2018 NRC Exam Page: 218 of 282 Question 76

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 219 of 282 Question 77

77 K/A Importance: 4.3/4.4 Points: 1.00 S77 Difficulty: 3.00 Level of Knowledge: Low Source: BANK 74251 The reactor has scrammed and the plant has been stabilized. The CRS has directed the scram to be RESET per 20.000.21, Reactor Scram.

What action must be taken if a Control Rod does not settle to the 00 position following reset of the scram?

A. Disarm the Control Rod per 23.106 Instrument Trip Sheets.

B. Verify rod position on the Rod Worth Minimizer.

C. Verify the Full-In light is lit on the Full Core Display.

D. Attempt to insert the Control Rod to 00 per 23.623 Reactor Manual Control System.

Answer: D ILO 2018 NRC Exam Page: 220 of 282 Question 77

Answer Explanation:

For control rods that did not settle to 00, the Reactor Scram AOP states to attempt to insert CR to 00 per 23.623.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of AOP 20.00.21, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (SCRAM and reset) and then taking action based on the requirements of AOP 20.000.21.

Distractor Explanation:

A. is plausible, but would be only be performed if manual insertion of the CR was unsuccessful and it is not a high friction rod.

B. is plausible but the RWM would only display number of rods out and not position.

C. is plausible, but would not tell the position of the control rod.

Reference Information:

20.000.21 Cond L Plant Procedures 20.000.21 NUREG 1123 KA Catalog Rev. 2 295006 AA2. Ability to determine and/or interpret the following as they apply to SCRAM :

295006 AA2.02 Control rod position 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage ILO 2013 Exam NRC Question Use (ILO 2018)

BANK Low SRO Associated objective(s):

AOP Objectives Cognitive Terminal Given a copy of 20.000.21, discuss licensed operator actions for a reactor scram.

ILO 2018 NRC Exam Page: 221 of 282 Question 77

78 K/A Importance: 3.3/3.1 Points: 1.00 S78 Difficulty: 3.00 Level of Knowledge: High Source: NEW 75107 The plant is operating at 100% power when an air leak on the Station Air Header downstream of the P5000-F402, Station Air to NIAS Iso Valve, occurs. Station and Control Air equipment status is:

• P5000-F440, Div 1 Control Air Iso Valve is CLOSED.

• P5000-F441, Div 2 Control Air Iso Valve is CLOSED.

• P5002-D001, Div 1 Control Air Compressor is RUNNING.

• P5000-D002, Div 2 Control Air Compressor is TRIPPED.

Following isolation of the leak on the air header, Station and Control Air pressures are:

• P50-R802, Station Air Header Pressure is 100 PSIG.

• P50-R870, IAS Header Pressure is 100 PSIG.

• Div 1 on P50-R801, NIAS Header Pressure Recorder, is 100 PSIG.

• Div 2 on P50-R801, NIAS Header Pressure Recorder, is 20 PSIG.

Which of the following will the CRS direct the CRLNO to perform from 20.129.01, Loss of Station and/or Control Air AOP?

A. Open P5000-F403, IAS to Div 2 NIAS Iso Vlv.

B. Open P5000-F441, Div 2 Control Air Iso Valve.

C. Cross-Tie NIAS, with Div 1 supplying, in accordance with 23.129, Station and Control Air System SOP.

D. Consult ARP 17D23, Div 2 Drywl Pneu Supply Iso Valves Closed for approximate time until MSIV isolation.

Answer: C ILO 2018 NRC Exam Page: 222 of 282 Question 78

Answer Explanation:

The candidate should interpret the conditions in the stem of the question as indicating a loss of Division 2 Non-Interruptible Air System (NIAS), or Division 2 safety related Instrument Air, pressure. Per AOP 20.129.01, with the Div 1 and Div 2 air headers intact, the candidate should determine that the correct course of action is to direct the NIAS headers to be cross-tied, with Division 1 supplying. This is per Condition E and F of the AOP.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of AOP 20.129.01, it is not related to immediate actions, and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (Instrument Air System pressure) and then directing action based on the requirements of AOP 20.129.01.

Distractor Explanation:

A. This distractor is plausible because it is also directed from Condition E of AOP 20.129.01 and, since IAS is available, the candidate could determine that opening of the P5000-F403 is warranted. However, per NOTE 1 of Condition E, with Div 2 NIAS pressure below 80 psig, the P5000-F403 will NOT open so this action is not correct and should not be directed by the CRS. Note: The keylock switch bypasses closure of the F403 on loss of power and low Station Air Header Pressure but it does NOT defeat closure of the F403 on low (<80 psig) Div 2 NIAS Header Pressure.

B. This distractor is plausible since re-opening of this valve is directed from Condition K of AOP 20.129.01 and the candidate could determine that, with the leak isolated and IAS header pressure restored to normal (100 PSIG), re-opening of the P5000-F441 is warranted in order to recover Div 2 NIAS. However, although the Low Station Air Pressure isolation (75 psig per Condition C of the AOP) for the P5000-F441 is clear, the valve also recieves an isolation on low Div 2 NIAS header pressure (sensed downstream of the P5000-F441 itself) at 75 PSIG so the valve is incapable of being opened until Div 2 NIAS header pressure is restored via another means, such as cross-tying NIAS headers.

D. This distractor is plausible because, with Div 2 NIAS Header pressure at 20 psig, Division 2 Drywell Pneumatics would have isolated, and 17D23 would be in, due to loss of air pressure to the pneumatic actuators for the T4901-F468 and F469, Div 2 Pneumatic Supply and Outboard Isolation Valves. The candidate could confuse Division 1 Drywell Pneumatics, which supplies the Inboard MSIVs with Nitrogen, with Division 2. This could lead the candidate to determine that Condition E of the AOP has a step similar to step D.1, which has the crew review ARP 8D66 when Division 1 Drywell Pneumatics isolates. This distractor is incorrect, however, because Division 2 Drywell Pneumatics does not supply nitrogen to the MSIVs and the AOP does not include this action.

Reference Information:

20.129.01 20.129.01 BASIS Plant Procedures 20.129.01 NUREG 1123 KA Catalog Rev. 2 295019 AA2. Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR :

295019 AA2.01 Instrument air system pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

High NEW SRO ILO 2018 NRC Exam Page: 223 of 282 Question 78

Associated objective(s):

AOP Objectives Cognitive Terminal Given a copy of 20.129.01, discuss licensed operator actions for a Loss of Station and/or Control Air.

ILO 2018 NRC Exam Page: 224 of 282 Question 78

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 225 of 282 Question 79

79 K/A Importance: 3.8/4.0 Points: 1.00 S79 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED 74247 The plant was operating at rated power when a partial loss of drywell cooling condition occurred.

• T48-R808, Primary Containment Pressure Rec is indicating 18 inches of water and is increasing at 0.1 inches every 2 minutes.

• T47-R803A, Average Drywell temperature is 137°F and rising 0.5°F every 10 minutes.

(1) What action(s) will the CRS direct first and (2) what is the basis for this action?

A. (1) Lower drywell pressure in accordance with 23.406, Primary Containment Nitrogen and Purge system using RBHVAC.

(2) Maintaining the expected initial conditions to ensure that, in the event of a DBA, equipment in the primary containment is able to perform its design function.

B. (1) Lower drywell temperature in accordance with EOP 29.100.01, Sheet 2, Primary Containment Control.

(2) Maintaining the expected initial conditions to ensure that, in the event of a DBA, equipment in the primary containment is able to perform its design function.

C. (1) Lower drywell pressure in accordance with 23.406, Primary Containment Nitrogen and Purge system using RBHVAC.

(2) Maintaining the expected initial conditions to ensure that the peak LOCA primary containment internal pressure does not exceed 62 psig.

D. (1) Lower drywell temperature in accordance with EOP 29.100.01, Sheet 2, Primary Containment Control.

(2) Maintaining the expected initial conditions to ensure that the peak LOCA primary containment internal pressure does not exceed 62 psig.

Answer: C ILO 2018 NRC Exam Page: 226 of 282 Question 79

Answer Explanation:

Drywell temperature and pressures are elevated, but they are not at EOP entry conditions. The correct action is to lower drywell pressure using either SBGTS or RBHVAC in accordance with 23.406. P&L 3.10.3 Directs Maintain pressure 5 to 19 inches of water as read on T48-R808, Primary Containment Pressure Rec 10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of SOP P&L and the basis behind the P&L, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions (approaching P&L limit) and then selecting a SOP with which to proceed.

Distractor Explanation:

(1) "Lower drywell temperature in accordance with EOP 29.100.01, Sheet 2, Primary Containment Control." is plausible because drywell temperature is high, but not high enough to enter the EOPs. Also, the high pressure condition was caused by a partial loss of drywell cooling.

(2) "Maintaining the expected initial conditions and ensures that so that In the event of a DBA, equipment in the primary containment is able to perform its design function." is the TS reason for maintaining DW temperture. While this answer is CORRECT when paired with distractor (1) for temperture, it is INCORRECT when paired with the answer for (1) pressure.

Reference Information:

TSB 3.1.6.4 23.406 P&Ls Plant Procedures 03D081 23.406 NUREG 1123 KA Catalog Rev. 2 295024 High Drywell Pressure.

G2.1.32 Ability to explain and apply system limits and precautions 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

High MODIFIED SRO Associated objective(s):

Nitrogen Purge, Inerting and Venting Performance Terminal Analyze Nitrogen Purge, Inerting and Venting response using control room indications and annunciators.

ILO 2018 NRC Exam Page: 227 of 282 Question 79

80 K/A Importance: 3.2/4.2 Points: 1.00 S80V3 Difficulty: 3.00 Level of Knowledge: High Source: NEW 75006 Suppression pool average temperature is > 111°F and rising. Per 3.6.2.1 Suppression Pool Average Temperature, when the suppression pool average temperature is > 121°F, a required action is to have reactor pressure less than 200 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and be in MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The reason for this action is:

A. To bring the plant to a MODE in which the LCO does not apply.

B. To preserve the heat absorption capability of the suppression pool.

C. To ensure that licensing bases initial conditions are met for the LCO.

D. Because the LCO is not met and an associated Required Action and Completion Time is not met and no other Condition applies.

Answer: A ILO 2018 NRC Exam Page: 228 of 282 Question 80

Answer Explanation:

Per TS Basis for 3.6.2.1 A limitation on the suppression pool average temperature is required to assure that the containment conditions assumed for the safety analyses are met. This limitation subsequently ensures that peak primary containment pressures and temperatures do not exceed maximum allowable values during a postulated DBA or any transient resulting in heat up of the suppression pool. The LCO requirements are:

a. Average temperature s 95°F with THERMAL POWER > 1% RATED THERMAL POWER (RTP), and no testing that adds heat to the suppression pool is being performed. This requirement ensures that licensing bases initial conditions are met.

b. Average temperature g 105 °F with THERMAL POWER > 1% RTP and testing that adds heat to the suppression pool is being performed. This required value ensures that the unit has testing flexibility, and was selected to provide margin below the 110°F limit at which reactor shutdown is required. When testing ends, the temperature must be restored to

• 95°F within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> according to Required Action A.2.

Therefore, the time that the temperature is > 95°F is short enough not to cause a significant increase in unit risk.

c. Average temperature

• 110°F with THERMAL POWER, 1% RTP. This requirement ensures that the unit will be shut down at > 110°F. The pool is designed to absorb decay heat and sensible heat but could be heated beyond design limits by the steam generated if the reactor is' not shut down.

E.1 and E.2 If suppression pool average temperature cannot be maintained at < 120°F, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the reactor pressure must be reduced to

< 200 psig within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and the plant must be brought to at least MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The allowed Completion Times is reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.

Continued addition of heat to the suppression pool with suppression pool temperature > 120°F could result in exceeding the design basis maximum allowable values for primary containment temperature or pressure. Furthermore, if a blowdown were to occur when the temperature was > 120°F, the maximum allowable bulk and local temperatures could be exceeded very quickly.

10 CFR 55.43(b)(2) SRO JUSTIFICATION This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of applying a TS and the basis for the actions. The answer is not related to <

1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action. The answer cannot be answered LCO/TRM information listed above-the-line and cannot be answered solely by knowing the TS Safety Limits. The SRO must assess plant conditions and then evaluate operability then choose the correct action and identify the correct basis.

Distractor Explanation:

B. "preserve the heat absorption capability of the suppression pool" incorrect because the temperature is high enough that there is not enough capacity left to perform the function of the suppression pool, and is plausible because it is a reason for an action required by this TS.

C. "ensures that licensing bases initial conditions are met for the LCO" incorrect because the action does not restore suppression pool ability to absorb energy, and is plausible because this is the basis for the LCO.

D. "To take T.S required actions because an LCO is not met and An associated Required Action and Completion Time is not met and no other Condition applies." is saying use LCO 3.03. This is incorrect because an LCO action applies. This is a logical distractor for TS basis.

Reference Information:

TS 3.6.2.1 and TSB 2.6.2.1 (not provided)

ILO 2018 NRC Exam Page: 229 of 282 Question 80

NUREG 1123 KA Catalog Rev. 2 295026 Suppression Pool High Water Temperature G2.2.25 Knowledge of bases in Technical Specifications for limiting conditions for operations and safety limits.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

NRC Question Use (ILO 2018)

High NEW NRC Early Review SRO Associated objective(s):

Technical Specifications for Licensed Operators Performance Enabler Given plant conditions that constitute non-compliance with any LCO, apply Technical Specifications and Bases to determine the applicable Condition(s), Required Action(s), and associated Completion Time(s).

ILO 2018 NRC Exam Page: 230 of 282 Question 80

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 231 of 282 Question 81

81 K/A Importance: 3.8/3.9 Points: 1.00 S81 Difficulty: 3.00 Level of Knowledge: High Source: BANK 74128 Following a transient, the following conditions exist:

• Drywell Temperature is 280ºF and stable.

• Drywell Pressure is 15 psig.

• Reactor Pressure is 20 psig.

• B21-R610 RPV CORE LEVEL RECORDER indication is oscillating between -20 inches and +30 inches

• All other RPV level indicators are off scale low or failed.

• ALL RHR Pumps are injecting.

Which one of the following actions is required?

A. Enter 29.100.01 Sheet 3, RPV Flooding, and inject until the RPV is flooded to the Main Steam Lines.

B. Open Turbine Bypass Valves and rapidly depressurize the reactor per 29.100.01 Sheet 1, RPV Control.

C. Enter 29.100.01 Sheet 3, Emergency Depressurization, and inject until RPV Pressure is > 64 psig above Torus Pressure.

D. Wait until RPV Water Level lowers to -43 inches and then Emergency Depressurize per 29.100.01 Sheet 3, Steam Cooling.

Answer: A ILO 2018 NRC Exam Page: 232 of 282 Question 81

Answer Explanation:

The candidate must determine that the RPV Saturation Temperature curve is exceeded with the given values of RPV Pressure and Drywell Temperature. With the Saturation Curve exceeded, and level indicators oscillating, RPV Water Level is unknown. With Level Unknown, it is required to Enter 29.100.01 Sheet 3, RPV Flooding, and inject until the RPV is flooded to the Main Steam Lines. (L-OR1) 10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOP directed RPV Flooding, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions and then selecting a section of the EOPs with which to proceed.

Distractor Explanation:

B. Is plausible; and would be true for other conditions where ED is anticipated.

C. Is plausible; >64 psig above Torus pressure with <4 SRVs open leads to use of Alternate Emergency Depressurization Systems.

D. Is plausible; and is required action for Steam Cooling.

Reference Information:

LP-OP-802-3003-0013C 29.100.01 SH3 29.100.01 SH1 Plant Procedures 29.100.01 SH 3 NUREG 1123 KA Catalog Rev. 2 295028 EA2. Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE :

295028 EA2.02 Reactor pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

BANK High SRO Associated objective(s):

EOP Objectives Cognitive Terminal Given a copy of 29.100.01 sheets 3 and 3A, discuss the operator's actions for Reactor Pressure Vessel Flooding (RF)

ILO 2018 NRC Exam Page: 233 of 282 Question 81

82 K/A Importance: 4.1*/4.5* Points: 1.00 S82 Difficulty: 3.00 Level of Knowledge: High Source: NEW 74126 A fuel failure has occurred and currently a Radiation Release in progress. The Offsite rad release rate is higher than the ALERT offsite release rate and less than the GENERAL EMERGENCY release rate. Use of HPCI is required to maintain Reactor Water Level. HPCI room temperature is 111°F and slowly rising.

Which of the following systems would the CRS direct to be isolated, per the Rad Release section of the EOPs, to mitigate the High Offsite Release Rate?

A. HPCI B. Off Gas System C. Main Steam Lines D. Standby Gas System Answer: C ILO 2018 NRC Exam Page: 234 of 282 Question 82

Answer Explanation:

per 29.100.01 SH 5 RR-2 Isolate primary systems that are discharging into areas outside the primary and secondary containment, except systems reqd by the EOPs.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOPs, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions and then applying the requirements of the EOPs.

Distractor Explanation Distractor are incorrect and plausible because:

A. HPCI could be a source that meets the requirements for isoloation however it is neccesary for the EOPs so it is incorrect.

B. Off Gas System is not a primary system, however it is a system that discharges to the enviroment.

D. Standby Gas is not a primary system, however it is a system that discharges to the enviroment.

Reference Information:

29.100.01 SH 5 Plant Procedures 29.100.01 SH 5 NUREG 1123 KA Catalog Rev. 2 295038 EA2. Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE :

295038 EA2.04 Source of off-site release 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

High NEW SRO Associated objective(s):

EOP Objectives Cognitive Terminal Given a copy of 29.100.01 sheet 5, discuss the operator's actions for Radioactivity Release Control (RR)

ILO 2018 NRC Exam Page: 235 of 282 Question 82

83 K/A Importance: 4.4/4.0 Points: 1.00 S83V2 Difficulty: 3.00 Level of Knowledge: High Source: NEW 75906 Due to plant condtions, the MODE Switch has been placed in SHUTDOWN. A SCRAM does NOT occur:

Based on this the CRS would observe the reactor operator __(1)__ both C7100-M606A MANUAL SCRAM TRIP A2 SWITCH and C7100-M606B MANUAL SCRAM TRIP B2 SWITCH located on the __(2)__ panel.

When FSQ1-8 is complete TRIP SYSTEM A/B indicate:

Based on this, the CRS will direct __(3)__

A. (1) Rotating (2) P602 (3) 29.ESP.10 DEFEAT OF ARI LOGIC TRIPS B. (1) Depressing (2) P603 (3) 29.ESP.09 DEFEAT OF RPS AUTOMATIC LOGIC TRIPS C. (1) Rotating (2) P602 (3) 29.ESP.09 DEFEAT OF RPS AUTOMATIC LOGIC TRIPS D. (1) Depressing (2) P603 (3) 29.ESP.10 DEFEAT OF ARI LOGIC TRIPS Answer: D ILO 2018 NRC Exam Page: 236 of 282 Question 83

Answer Explanation:

Per ODE 10 If all control rods are not fully inserted after the Reactor Mode Switch has been placed in Shutdown, the P603 will depress the RPS Manual Scram pushbuttons and monitor for proper response. -

This action should hace made the 8 blue RPS lights turn off.

Per 29.ESP.03 Step 4.1 If the 8 blue RPS lights are off defeat RPS logic trips in accordance with 29.ESP.09, Defeat of RPS Automatic Logic Trips. Based on condtions 29.ESP.10 is required.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of actions required by the EOPs and sub procedures (ESPs), it is not related to immediate actions and the entry conditions are not relevant or leading to answer. The SRO must assess the plant conditions (ATWS) and then diagnose the status of RPS per 29.ESP.03 and then ensure 29.ESP.09 is used to mitigate / recover.

Distractor Explanation:

(1) "Rotate" and (2) "P602" is incorrect and plausible because the label for these push buttons is "SWITCH" and the P602 is in the front arc of Main Control Room.

(3) 29.ESP.03 allows for 29ESP.09, based condtions this is incorrect.

Reference Information:

ODE 10 Pg 8 29.ESP.03 EOPs Plant Procedures 29.ESP.03 NUREG 1123 KA Catalog Rev. 2 295015 Incomplete SCRAM G2.1.30 Ability to locate and operate components, including local controls 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

High NEW NRC Early Review SRO Associated objective(s):

Emergency Support Procedures (ESP)

Cognitive Terminal Given a copy of 29.ESP.03, discuss Licensed Operator actions for Alternate Control Rod Insertion ILO 2018 NRC Exam Page: 237 of 282 Question 83

84 K/A Importance: 3.5*/3.9* Points: 1.00 S84 Difficulty: 3.00 Level of Knowledge: High Source: BANK 74047 The plant is operating when a small break LOCA and ATWS occurs. ATWS conditions continue for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> before the crew is able to shutdown the reactor (all rods are full in). The following plant conditions are now present:

• RPV water level is 10" above TAF and steady.

• SBFW is the only injection source.

• Reactor Pressure is 600 psig and rising.

• MSIVs are closed.

• Drywell pressure is 12 psig and steady.

• Drywell Temperature is 280°F and steady.

• Suppression pool temperature is 152°F and steady.

• Torus water level is 16 and slowly rising, and all efforts to lower level are unsuccessful.

If the current trend in primary containment water level persists, what action is required and why?

A. Emergency Depressurize to protect the integrity of the SRV tailpipes.

B. Emergency Depressurize to preserve pressure suppression capability.

C. Reduce reactor pressure (<90°F/hr) to protect the integrity of the SRV tailpipes.

D. Reduce reactor pressure (<90°F/hr) to preserve pressure suppression capability.

Answer: A ILO 2018 NRC Exam Page: 238 of 282 Question 84

Answer Explanation:

Per EPG (B-17-80) the SRVTPLL (STPLL in EPG) bases is:

The SRV Tail Pipe Level Limit (STPLL) is the lesser of:

• The Maximum Pressure Suppression Primary Containment Water Level.

• The highest suppression pool water level at which opening an SRV will not result in exceeding the code allowable stresses in the SRV tail pipe, tail pipe supports, quencher, or quencher supports.

The STPLL is a function of RPV pressure. SRV operation with suppression pool water level above the STPLL could damage the SRV discharge lines. This, in turn, could lead to containment failure from direct pressurization and damage to equipment inside the containment (ECCS piping, RPV water level instrument runs, wetwell-to-dry well vacuum breakers, etc.) from pipe-whip and jet-impingement loads.

Based on the stem of the question Operation is currently in the unsafe area of the SRVTPLL. Efforts to lower level have so far been unsuccessful. ED is required to protect the SRV tail pipes per TWL-11. This requires the SRO to execute P-OR1-3, ED is REQ'D 10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of actions required by the EOPs, it is not related to immediate actions and the entry conditions are not relevant or leading to answer. The SRO must assess the plant conditions (SRVTPLL) and then selecting the step of the EOP to mitigate / recover.

Distractor Explanation:

B. Is incorrect because the reason ED is required is to protect the SRV tail pipe. This answer is plausible based on not understanding the bases of the SRVTPLL C. Is incorrect because an ED is required and taking P-OR1-3 requires exiting pressure leg. This answer is plausible based on not assessing the need to ED.

D. Is incorrect because an ED is required and taking P-OR1-3requires exiting pressure leg. This answer is plausible based on not assessing the need to ED.

Reference Information:

BWROG EPG/SAGs Appendix B Bases B-17-80 Plant Procedures 29.100.01 SH 1A BWROG EPG App B NUREG 1123 KA Catalog Rev. 2 295029 EA2.01 Suppression pool water level 295029 EK3. Knowledge of the reasons for the following responses as they apply to HIGH SUPPRESSION POOL WATER LEVEL :

295029 EK3.01 Emergency depressurization 295030 EA2. Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL :

10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage LOR 2009 Exam LOR 2015 Exam NRC Question Use (LOR 2017)

SRO ONLY NRC Question Use (ILO 2018)

BANK High SRO ILO 2018 NRC Exam Page: 239 of 282 Question 84

Associated objective(s):

Cautions, Curves, and Calculations Cognitive Enabler Discuss the definition of and reason for the shape of the following graphs/curves:

a. Boron Injection Initiation Temperature Curve

b. Core Spray Net Positive Suction Head (NPSH) Limit

c. Core Spray Vortex Limit

d. Heat Capacity Limit Deleted

f. Deleted

g. RHR Low Pressure Coolant Injection (LPCI) NPSH Limit

h. RHR LPCI Vortex Limit

i. RPV Saturation Temperature

j. SRV Tail Pipe Level Limit

k. Drywell Spray Initiation Limit

l. Pressure Suppression Pressure

m. Primary Containment Pressure Limit

n. HPCI NPSH Limit

o. RCIC NPSH Limit EOP Objectives Cognitive Terminal Given a copy of 29.100.01 sheet 1A, discuss the operator's actions for Anticipated Transient Without Scram (ATWS) to control Reactor Pressure Vessel pressure (FSP)

ILO 2018 NRC Exam Page: 240 of 282 Question 84

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 241 of 282 Question 85

85 K/A Importance: 3.8/3.9 Points: 1.00 S85 Difficulty: 2.00 Level of Knowledge: Low Source: BANK 74011 During the execution of 29.100.01 Sheet 5, Secondary Containment and Radiation Release, the operability of equipment required to perform a safe shutdown is assured by Emergency Depressurization for which of the following conditions?

A. Radiation Level in any ONE AREA EXCEEDS the MAX NORMAL value.

B. Radiation Levels in MORE THAN ONE AREA EXCEED the MAX SAFE value.

C. Radiation Levels in MORE THAN ONE AREA EXCEED the MAX NORMAL value.

D. Radiation Level in any ONE AREA is CONFIRMED by a MAX SAFE Area Temperature OR MAX SAFE Water Level in the SAME AREA.

Answer: B ILO 2018 NRC Exam Page: 242 of 282 Question 85

Answer Explanation:

Emergency Depressurization is required when the Radiation Levels in MORE THAN ONE AREA exceed the MAX SAFE value.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOP directed Emergency Depressurization, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions and then applying the requirements of the EOPs.

Distractor Explanation:

A is plausible; and is an entry condition for 29.100.01 Sheet 5.

C is plausible; does NOT require Emergency Depressurization.

D is plausible; does NOT require Emergency Depressurization.

Reference Information:

EOP 29.100.01 Sheet 5 Objective Link: LP-OP-802-3005-0009 Plant Procedures 29.100.01 SH 1 NUREG 1123 KA Catalog Rev. 2 295033 EA2. Ability to determine and/or interpret the following as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS :

295033 EA2.01 Area radiation levels 295033 EA2.02 Equipment operability 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

BANK Low SRO Associated objective(s):

Emergency Support Procedures (ESP)

Cognitive Enabler Given a copy of 29.100.01 SH1/1A, discuss Licensed Operator actions to Rapidly depressurize the Reactor Pressure Vessel using the Main Steam Line Drains ILO 2018 NRC Exam Page: 243 of 282 Question 85

86 K/A Importance: 3.4/3.5 Points: 1.00 S86VB Difficulty: 3.50 Level of Knowledge: High Source: BANK 76086 The plant just entered Mode 4 in preparation for a Refueling Outage. Div 1 RHR aligned for SDC.

A leak develops in the RWCU system which isolates on high differential flow. The following alarms and indications are observed:

• 1D23, DIV I RHR PUMP A/C MOTOR TRIPPED

• 3D79 REAC VESSEL WATER LEVEL L3

• RPV water level is 170 inches and steady.

• RPV temperature is 150°F and rising slowly.

As CRS, (1) what procedure(s) would be selected to mitigate this event and (2) what actions are the minimum required to be performed to allow restoration of Div 1 RHR to Shutdown Cooling?

A. (1) 20.205.01 LOSS OF SHUTDOWN COOLING (2) Reset NSSS logic, and depress E1150-F015A Seal In Reset pushbutton.

B. (1) 20.205.01 LOSS OF SHUTDOWN COOLING, 29.100.01 SH 5 SEC CONT/RAD RELEASE (2) Reset NSSS logic, and depress E1150-F015A Seal In Reset pushbutton.

C. (1) 20.205.01 LOSS OF SHUTDOWN COOLING, 29.100.01SH 1 RPV CONTROL (2) Raise level above 220, reset NSSS logic, and depress E1150-F015A Seal In Reset pushbutton.

D. (1) 20.205.01 LOSS OF SHUTDOWN COOLING, 29.100.01 SH 1 RPV CONTROL, 29.100.01 SH 5 SEC CONT/RAD RELEASE (2) Raise level above 220, reset NSSS logic, and depress E1150-F015A Seal In Reset pushbutton.

Answer: C ILO 2018 NRC Exam Page: 244 of 282 Question 86

Answer Explanation:

Plant conditions do not meet the entry conditions for Sheet 5, but meet those in 20.205.01 and Sheet 1 because of RPV Level 3. Actions are contained in condition C of 20.205.01.

A. is incorrect because you meet entry conditions for sheet 1. This directs you to raise level as does the AOP.

B. is incorrect because you do not meet entry conditions for sheet 5, you meet entry conditions for sheet 1.

This directs you to raise level as does the AOP.

D. is incorrect because you do not meet entry conditions for sheet 5.

Reference:

20.205.01, page 5 Plant Procedures 20.205.01 29.100.01 SH 1 29.100.01 SH 5 NUREG 1123 KA Catalog Rev. 2 205000 A2. Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLINGMODE) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

205000 A2.06 SDC/RHR pump trips NRC Exam Usage ILO 2012 Exam NRC Question Use (ILO 2018)

BANK High SRO Associated objective(s):

Cycle 12-1 Objectives Performance Objectives Performance Terminal Upon recognizing the entry conditions for an inadvertent loss of RHR shutdown cooling operation, operators respond to control critical parameters within expected limits IAW approved plant procedures within validated completion times ILO 2018 NRC Exam Page: 245 of 282 Question 86

87 K/A Importance: 3.4/3.6 Points: 1.00 S87V2 Difficulty: 3.00 Level of Knowledge: High Source: NEW 75907 The plant has scrammed.

• The Drywell Pneumatics supply header is depressurized due to a rupture.

• Conditions exist that require Emergency Depressurization per the EOPs.

• Drywell pressure is 5 psig.

• Torus water level is 2 inches.

What action will the CRS direct?

A. Open 5 ADS SRVs.

B. Open 5 Divison 2 SRVs C. Open Main Turbine bypass valves, ignoring cooldown rates.

D. Blowdown using RWCU, install defeats as necessary from 29.ESP.18 Defeat of RWCU Isolations.

Answer: A ILO 2018 NRC Exam Page: 246 of 282 Question 87

Answer Explanation:

Per EOP Open 5 SRVs, ADS preferred 10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOP directed Emergency Depressurization, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions and then applying the requirements of the EOPs.

Distracter Explanation:

Distracters are incorrect and plausible because B. All Drywell Pneumatics is depressurized due to a rupture, therefore the misconception is that Division 1 SRVs not available. However per EOP ADS SRV are used.

C. Open Main Turbine bypass valves, ignoring cooldown rates is an override for anticipate ED and a possible option if SRVs are not available per table 18 of EOPs. It is not an authorized option because 5 ADS SRVs are available D. RWCU blowdown is only used if SRVs are not available and is per table 18 of EOPs.

Reference Information:

29.100.01.SH3 ED-4 M-5007 Plant Procedures 29.100.01 SH 3 NUREG 1123 KA Catalog Rev. 2 218000 A2. Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

218000 A2.03 Loss of air supply to ADS valves: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

High NEW NRC Early Review SRO Associated objective(s):

EOP Objectives Cognitive Terminal Given a copy of 29.100.01 sheet 3, discuss the operator's actions for Reactor Pressure Vessel Emergency Depressurization (ED)

ILO 2018 NRC Exam Page: 247 of 282 Question 87

88 K/A Importance: 3.4/4.7 Points: 1.00 S88 Difficulty: 4.00 Level of Knowledge: High Source: NEW 73909 Review the table of as found SRV lift pressures. The Plant is in MODE 1.

SRV # SRV # SRV #

A 1080 D 1134 E 1140 B 1122 F 1182 H 1193 C 1075 L 1080 J 1158 G 1099 M 1165 P 1116 K 1120 N 1155 R 1139 Based on T.S. 3.4.3 how many SRV setpoint(s) must be adjusted to restore the minimum number of required OPERABLE SRVs?

A. 0 B. 1 C. 2 D. 3 Answer: B ILO 2018 NRC Exam Page: 248 of 282 Question 88

Answer Explanation:

ILO 2018 NRC Exam Page: 249 of 282 Question 88

Per SR 3.4.3.1 Per BASIS for SR 3.4.3.1 This Surveillance requires that the required SRVs will open at the pressures assumed in the safety analysis of Reference 1. The demonstration of the SRV safe lift settings must be performed during shutdown, since this is a bench test, to be done in accordance with the Inservice Testing Program. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures. The SRV setpoint is +/- 3% for OPERABILITY, however, the valves are reset to +/- 1% during the Surveillance to allow for drift. The SR gives set pressures for all 15 SRVs installed. However, since only 11 SRVs are required, the SR is met if 11 SRVs are set properly.

SRV DATA:

SRV AS Required IN RANGE FOUND A 1080 1135 B 1122 1135 1135 C 1075 1135 G 1099 1135 K 1120 1135 1135 D 1134 1145 1145 F 1182 1145 1155 L 1080 1145 M 1165 1145 1145 N 1155 1145 1145 E 1140 1155 1155 H 1193 1155 J 1158 1155 1155 P 1116 1155 1145 R 1139 1155 1155

@1135 2 ILO 2018 NRC Exam Page: 250 of 282 Question 88

@1145 4

@1155 4 Total 10 Therefore 1 must MUST be adjusted 10 CFR 55.43(b)(2) SRO JUSTIFICATION This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of applying SR 3.4.3.1. The answer is related to < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action however it cannot be soley answered with this information because OPERABILITY must first be determined using a SR. The answer cannot be answered LCO/TRM information listed above-the-line and cannot be answered by solely by knowing the TS Safety Limits. The SRO must assessing plant conditions and then evaluate operability based on SR 3.4.3.1 and then choose the correct action.

Distractor Explanation:

Because 2 of the as found values are not in the accepable band, but still in an acceptable range for tech spec, and at least 1 must be adjusted a max of 3 is plausible.

Reference Information:

TS for 3.4.3 and SR 3.4.3.1 (provided)

NUREG 1123 KA Catalog Rev. 2 239002 SRVs G2.2.40 Ability to apply technical specifications for a system 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

NRC Question Use (ILO 2018)

High NEW NRC Early Review SRO Associated objective(s):

Technical Specifications for Licensed Operators Performance Enabler Given plant conditions that constitute non-compliance with any LCO, apply Technical Specifications and Bases to determine the applicable Condition(s), Required Action(s), and associated Completion Time(s).

ILO 2018 NRC Exam Page: 251 of 282 Question 88

89 K/A Importance: 4.4/4.7 Points: 1.00 S89V2 Difficulty: 4.00 Level of Knowledge: High Source: MODIFIED 75008 The plant is operating at 100% power. A review of the Quarterly Battery Check surveillance, for Division II shows the following results:

• Electrolyte Level 1/4" above maximum

• Lowest Specific Gravity 1.192

• Lowest Float Voltage 2.06 (3 cells)

Which of the following action(s), if any would be required for the given conditions?

A. T.S. 3.8.6 ACTION B.1 ONLY B. T.S. 3.8.6 ACTION A.1, A.2 and A.3 ONLY C. T.S. 3.8.6 ACTION A.1, A.2, A.3 AND T.S. 3.8.6 ACTION B.1 ONLY D. T.S. 3.8.6 ACTION A.1, A.2 and A.3 AND T.S. 3.8.6 ACTION B.1 AND T.S. 3.8.4 ACTION B.1 Answer: D ILO 2018 NRC Exam Page: 252 of 282 Question 89

Answer Explanation:

TS 3.8.6 Table 3.8.6.1 Category C states float voltage must be >2.07V for each connected cell. This requires entry into TS 3.8.6 Condition B to declare associated battery inoperable. This also requires entry into TS 3.8.4 Condition B for a DC electrical subsystem inoperable with a completion to operable status of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> 10 CFR 55.43(b)(2) SRO JUSTIFICATION This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of applying the SRs in 3.8.6 (Table 3.8.6-1). The answer is related to < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action however it cannot be soley answered with this information because OPERABILITY must first be determined using a SR. The answer cannot be answered LCO/TRM information listed above-the-line and cannot be answered by solely by knowing the TS Safety Limits. The SRO must assessing plant conditions and then evaluate operability based on TS 3.8.6 and TS 3.8.4. and then choose the correct action Distractor Explanation:

A. "T.S. 3.8.6 Cond B.1 ONLY" is incorrect because of the ONLY (see answer explanation). This distractor is plausible if the examinee reviews T.S. 3.8.4 and finds the DC subsystem to be OPERABLE.

B. "T.S. 3.8.6 Cond A.1, A.2 and A.3 ONLY" is incorrect because of answer explanation. This distractor is plausible if the examinee reviews T.S 3.8.6 can find the battery cells to be on meeting Category A or B limits.

C. See B & C Reference Information:

TS & TSB for 3.8.6 and 3.8.4 NUREG 1123 KA Catalog Rev. 2 263000 DC Electrical Distribution G2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation G2.2.40 Ability to apply technical specifications for a system Technical Specifications 3.8.4 DC Sources Operating 3.8.6 Battery Cell Parameters 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

NRC Question Use (ILO 2018)

High MODIFIED SRO Associated objective(s):

Technical Specifications for Licensed Operators Performance Enabler Given plant conditions that constitute non-compliance with any LCO, apply Technical Specifications and Bases to determine the applicable Condition(s), Required Action(s), and associated Completion Time(s).

ILO 2018 NRC Exam Page: 253 of 282 Question 89

90 K/A Importance: 4.2/4.2 Points: 1.00 S90V2 Difficulty: 2.00 Level of Knowledge: High Source: NEW 75127 The plant is at 100% power when a report about the material condition of EDG 14 requires it be declared INOP. A Nuclear Operator is dispatched to place EDG 14 in Maintenance Pullout. The following alarms are in after EDG 14 is reported in Maintenance Pullout.

(1) Are these alarms consistent with the expected plant conditions after placing EDG 14 in Maintenance Pullout?

(2) What is the most limiting time to restore OPERABILITY per TS 3.8.1?

A. (1) No (2) 14 Days B. (1) No (2) 72 Hours C. (1) Yes (2) 72 Hours D. (1) Yes (2) 14 Days Answer: B ILO 2018 NRC Exam Page: 254 of 282 Question 90

Answer Explanation:

Per 10D34 and 10D32, EDG 13 is in MPO(Maintenance Pullout).

The Nuclear Operator put the wrong EDG in MPO. This means BOTH EDG 13 and 14 are INOP Per TS 3.8.1 B.4 - 72 hrs to restore one EDG.

10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of T.S and how to apply actions. The answer is NOT related immediate actions. The examinee must determine operability first to then apply the correct actions. The answer is not related to TS Safety Limits. The answer is relevant to the entry conditions but is not leading to the answer. While information found above the double line is needed, the correct understanding of the bases for the condition is required. The answer to this questions is based on assessing plant conditions (Wrong EDG in MPO) and evaluating conditions and then applying information from the T.S Bases.

Distractor Explanation "Yes" is incorrect and plausible if the candidate cannot verify that the alarms are consistent with the plant conditions.

"14" days is incorrect and plausible because it is the answer for only 1 EDG INOP Reference Information:

10D34 and 10D32 TS/B 3.8.1 (Provided)

NUREG 1123 KA Catalog Rev. 2 264000 Emergency Generators (Diesel/Jet)

G2.4.46 Ability to verify that the alarms are consistent with the plant conditions 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

High NEW SRO Associated objective(s):

Technical Specifications for Licensed Operators Performance Enabler Given plant conditions that constitute non-compliance with any LCO, apply Technical Specifications and Bases to determine the applicable Condition(s), Required Action(s), and associated Completion Time(s).

ILO 2018 NRC Exam Page: 255 of 282 Question 90

91 K/A Importance: 3.4/3.4 Points: 1.00 S91 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED 73107 Reactor power at 74% when the B31005-F031A (A RR Pump Discharge Valve) open position switch failed causing a loss of its full open position indication.

(1) What is the expected plant response and (2) what shall the CRS direct?

A. (1) DCS Logic will run South RR MG Set back to the Limiter #1 setting.

(2) Direct LNO to:

• Take local manual control of North RR MG Set.

• Coordinate with the P603 opertor to match recirc flows.

B. (1) DCS Logic will run South RR MG Set back to the Limiter #4 setting.

(2) Direct P603 to:

• Verify Reactor Power < 66.1% or insert the Cram Array to lower Reactor Power to < 66.1%

• Increase core monitoring for instability.

C. (1) DCS Logic will run South RR MG Set back to the Limiter #1 setting.

(2) Direct P603 to:

• Verify Reactor Power < 66.1% or insert the Cram Array to lower Reactor Power to < 66.1%

• Increase core monitoring for instability.

D. (1) DCS Logic will run South RR MG Set back to the Limiter #4 setting.

(2) Direct LNO to:

• Take local manual control of North RR MG Set.

• Coordinate with the P603 opertor to match recirc flows.

Answer: B ILO 2018 NRC Exam Page: 256 of 282 Question 91

Answer Explanation:

RR Limiter 4 is actuated by the North (South) RR MG Set drive motor or North (South) RR MG Set generator field breakers opening. The A RR MG set breaker will trip on interlock because of the discharge valve not being full open. Once the logic is cleared (fixed), the MG set can be restarted, and the limiter must be manually reset.

Distractor Explanation:

Distractor (1) Plausible because the examinee may not understand that the MG tripped off. RR Limiter 1 is actuated when B3105-F031A (B), N (S) RR Pump Discharge Vlv, is not fully open or total Feedwater flow is less than 20% of original rated value (approximately 3.15 Million lb./hr). It is however incorrect because this distractor is the South RR pump.

Distractor (2) Plausible and incorrect because the examinee may not understand that the MG tripped off.

Based on this, matching flows is required by plant procedures and TS and therefore would be a priority for the CRS.

Reference Information:

23.138.01 Section 1.0 (pg 4-12) 20.138.01 Condition A (pg 3)

Plant Procedures 20.138.01 23.138.01 NUREG 1123 KA Catalog Rev. 2 202002 A2. Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

202002 A2.01 Recirculation pump trip 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

High MODIFIED SRO Associated objective(s):

Cycle 15-3 Objectives Cognitive Terminal Given a copy of 20.138.01, discuss licensed operator actions for a Recirculation Pump Trip ILO 2018 NRC Exam Page: 257 of 282 Question 91

92 K/A Importance: 3.6/4.6 Points: 1.00 S92 Difficulty: 3.00 Level of Knowledge: High Source: NEW 73166 All APRMs indicate 31% power. An interior control rod is currently selected.

Per Tech Specs which action, if any, is required?

Action Completion Time A. No action required.

B. Restore RBM channel A to OPERABLE status. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

C. Restore RBM channel B to OPERABLE status. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Place one RBM channel in trip. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

ILO 2018 NRC Exam Page: 258 of 282 Question 92

Answer: B ILO 2018 NRC Exam Page: 259 of 282 Question 92

Answer Explanation:

Per SR 3.3.2.1.5 Verify the RBM is not bypassed when THERMAL POWER is > 30% RTP.

Per TS 3.3.2.1 Condition A Restore RBM channel to OPERABLE status. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

10 CFR 55.43(b)(2) SRO JUSTIFICATION This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of applying SR 3.3.2.1.5. The answer is related to < 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> TS/TRM Action however it cannot be solely answered with this information because OPERABILITY must first be determined using a SR. The answer cannot be answered LCO/TRM information listed above-the-line and cannot be answered by solely by knowing the TS Safety Limits. The SRO must assessing plant conditions and then evaluate operability based on SR 3.3.2.1.5. and then choose the correct action.

Distractor Explanation:

Distractors are incorrect and plausible based on answer explanation.

"No action required." is incorrect per TS 3.3.2.1 and plausible if candidate does not understand status of the RBM A "Restore RBM channel B to OPERABLE status. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." is incorrect and because RBM is OPERABLE and plausible if the candidate does not understand status of the RBM A "Place one RBM channel in trip. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />." is incorrect and plausible because it is the correct action from T.S Reference Information:

TS 3.3.2.1 & TSB 3.3.2.1 NUREG 1123 KA Catalog Rev. 2 215002 RBM System G2.2.37 Ability to determine operability and/or availability of safety related equipment Technical Specifications 3.3.2.1 Control Rod Block Instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

NRC Question Use (ILO 2018)

High NEW NRC Early Review SRO Associated objective(s):

Technical Specifications for Licensed Operators Performance Enabler Given plant conditions that constitute non-compliance with any LCO, apply Technical Specifications and Bases to determine the applicable Condition(s), Required Action(s), and associated Completion Time(s).

ILO 2018 NRC Exam Page: 260 of 282 Question 92

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 261 of 282 Question 93

93 K/A Importance: 3.8/3.9 Points: 1.00 S93 Difficulty: 4.00 Level of Knowledge: High Source: NEW 73788 The plant is operating at 100% power with the following:

The inboard MSIV for Main Steam Line A slow closes and plant indications are:

(1) What is the total steam flow going to the 52" manifold and (2) based on the MSIV configuration, the CRS will direct a change of reactor power to?

A. (1) ~14.98 E 06 LB/HR (2) 90%

B. (1) ~12.75 E 06 LB/HR (2) 90%

C. (1) ~12.75 E 06 LB/HR (2) 59%

D. (1) ~14.98 E 06 LB/HR (2) 59%

ILO 2018 NRC Exam Page: 262 of 282 Question 93

Answer: D ILO 2018 NRC Exam Page: 263 of 282 Question 93

Answer Explanation:

A SLOW closure of a MSIV at power will result in a high reactor pressure. 3D168 will Alarm. Because the pressure regulator will maintain the same pressure in the 52" manifold, the Main Turbine continues with the same power output. This means that the same mass flow rate of steam is flowing through the turbine does not change To achieve this with 3 mainsteam lines pressure in the reactor vessel would have to rise to make up for the increase in steam velocity/mass flow rate going down 3 lines vice 4.

In this case the flow indications have a max value of 4.25 E 06, so they are indication high off scale.

Per 23.137 Reactor power must be less than 60% for a main steam line to be isolated.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of operating limits of the Main Steam Lines, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions and then applying the requirements of the SOPs.

Distractor Explanation:

~12.75 E 06 LB/HR is incorrect based on answer explaination. It is however the INDICATED flow based on the flow indication provided.

90% power is based on being lower than 91.5% power which is required per 23.109 Section 7.1 for closing off on a line to the High Pressure Turbine. However the answer is for a Main Steam line so this in incorrect.

Reference Information:

23.137 (Pg 15)

Plant Procedures 23.137 NUREG 1123 KA Catalog Rev. 2 239001 A2. Ability to (a) predict the impacts of the following on the MAIN AND REHEAT STEAM SYSTEM ; and (b) based on those predictions, use procedures to correct, control,or mitigate the consequences of those abnormal conditions or operations:

239001 A2.10 Closure of one or more MSIV's at power 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

High NEW NRC Early Review SRO Associated objective(s):

Nuclear Boiler System (B1100, B2100, B2103, B2104, N1100 & N3017)

Cognitive Terminal Given various controls and indications for Nuclear Boiler System operations, analyze that indication to determine proper plant response in accordance with approved plant procedures.

ILO 2018 NRC Exam Page: 264 of 282 Question 93

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 265 of 282 Question 94

94 K/A Importance: 3.9/4.2 Points: 1.00 S94 Difficulty: 2.00 Level of Knowledge: High Source: BANK 73106 Following an Emergency Depressurization under ATWS conditions the following conditions exist:

• RPV Water Level indications are UPSCALE due to High Drywell and Reactor Building Temperatures.

• Injection has been Terminated and Prevented.

• 4 Safety Relief Valves are OPEN.

• Reactor Pressure is 300 psig.

• Reactor Power is 20%.

Under these conditions, which one of the following describes the status of Core Cooling?

Adequate Core Cooling is:

A. ASSURED by core submergence.

B. ASSURED by sufficient steam flow through OPEN Safety Relief Valves.

C. NOT ASSURED because injection is insufficient to cool the core.

D. NOT ASSURED because an inadequate number of Safety Relief Valves are OPEN.

Answer: B ILO 2018 NRC Exam Page: 266 of 282 Question 94

Answer Explanation:

Per EOP/EPG, 29.100.01, Sheet 3A (provided to the candidates) with Reactor Pressure above 290 psig with 4 Safety Relief Valves OPEN, Adequate Core Cooling is ASSURED. This information is the basis for the Minimum Steam Cooling Pressure (MSCP), which is found on Table 8.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOP Adequate Core Cooling conditions, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions and then applying the requirements of the EOPs and knowing the bases of the EOPs/EPGs.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Would be true if UPSCALE RPV Water Level Instruments were valid.

C. If submergence were the only acceptable method of assuring Adequate Core Cooling, this would be correct. Terminated injection, long term, will not provide Adequate Core Cooling at 20% power. Long term issues are not addressed, the question pertains to Under these conditions.

D. Would be true if Reactor Pressure were under 290 psig.

Reference Information:

BWROG EPGs/SAGs, Appendix B (EOP) 29.100.01 SH 3A (provided)

Plant Procedures 29.100.01 SH 3 NUREG 1123 KA Catalog Rev. 2 216000 Nuclear Boiler System G2.1.25 Ability to interpret station reference materials such as graphs, curves, tables, etc.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

BANK High NRC Early Review SRO Associated objective(s):

EOP Objectives Cognitive Terminal Given a copy of 29.100.01 sheet 3, discuss the operator's actions for Steam Cooling (STC)

ILO 2018 NRC Exam Page: 267 of 282 Question 94

95 K/A Importance: 4.3/4.6 Points: 1.00 S95 Difficulty: 2.00 Level of Knowledge: Low Source: BANK 73009 The plant is operating at 100%. The SM is reviewing the Plan of the Day (POD) for the upcoming week. Which one of the following evolutions would require coverage from a Reactivity Management SRO (RMSRO) who is not the STA/IA?

A. HPCI surveillance.

B. Turbine Valve Testing surveillance.

C. Power change required for CRD Operability surveillance.

D. Down power to support removal of Reactor Feed Pump from service.

Answer: D ILO 2018 NRC Exam Page: 268 of 282 Question 95

Answer Explanation:

Per MOP-19, Large reactivity changes requiring a Reactivity Management Senior Reactor Operator (RMSRO) are planned changes to Control Rod position or Recirculation pump speed. MOP-19 then states that the RMSRO role is normally fulfilled by the on shift STA/IA with an active SRO license. MOP-19 goes on to list exceptions to the use of the STA/IA as the RMSRO. The candidate should recognize that the correct answer will involve a load drop of greater than 30% and restoration with major BOP equipment manipulations as being an exception to the use of the on shift STA/IA and therefore conclude that another individual will have to fulfill the RMSRO role.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of Reactivity changes requiring a Reactivity Management Senior Reactor Operator (RMSRO), it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing a planned reactivity change and implementating a station administrative procedure correctly.

Distracter Explanation:

A/B/C are all incorrect because they are examples of Minor Reactivity Adjustments, generally defined to be less than 10%. They are all plausible because they are listed in MOP19-100 as reactivity adjustments and the candidate could incorrectly recall the list as requiring the use of a RMSRO other than the on-shift STA/IA. However, the examples identified in MOP19-100, section 5.4.2, will either not require a RMSRO or will allow the RMSRO function to be fulfilled by the on-shift STA/IA.

Reference Information:

MOP19, Section 4.3 MOP19-100, Section 5.4.2.

Plant Procedures MOP19 ODE-01 Reactivity Management NUREG 1123 KA Catalog Rev. 2 G2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management NRC Question Use (ILO 2018)

BANK Low SRO Associated objective(s):

Conservative Reactivity Management Cognitive Terminal Discuss Detroit Edison expectations on Conservative Reactivity Management as outlined in MOP19.

ILO 2018 NRC Exam Page: 269 of 282 Question 95

96 K/A Importance: 2.3/3.4 Points: 1.00 S96V2 Difficulty: 3.00 Level of Knowledge: Low Source: NEW 76026 A surveillance testing work order and associated I&C surveillance procedure are being reviewed by a Shift's Field Support Supervisor (FSS), per MWC03 Surveillance/Performance Package Control. This review shall include, as a minimum, a review of:

A. 1. Plant configuration and mode

2. Required prerequisites

3. Impact statement (if required)

B. 1. Qualification of test personnel.

2. Required prerequisites

3. Impact statement (if required)

C. 1. Plant configuration and mode

2. Required prerequisites

3. Qualification of test personnel.

D. 1. Plant configuration and mode

2. Qualification of test personnel.

3. Impact statement (if required)

Answer: A ILO 2018 NRC Exam Page: 270 of 282 Question 96

Answer Explanation:

Per MWC03 4.2.5 Review the surveillance procedure to ensure that it can be performed under the existing plant configuration without any adverse impact on the plant. This review shall include, as a minimum:

1.Plant configuration and mode 2.Required prerequisites 3.Impact statement (if required) 10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of maintenance work order requirements from MWC03, it is not related to immediate actions, and the entry conditions are not relevant or leading to the answer. To answer this question, the examinee must know when to implement attachments and appendices, including how to coordinate these items with procedure steps. In this case MWC03 and surveillance testing work order review requirements.

Distractor Explanation:

Work hours and number of test personnel to complete is incorrect and plausible because, while it is reviewed, this review is done by planning. Work hour limits are a valid item that has federal limits.

Reference Information:

MWC03 4.2.5 (pg 11)

Plant Procedures MWC03 NUREG 1123 KA Catalog Rev. 2 G2.2.19 Knowledge of maintenance work order requirements 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILO 2018)

Low NEW SRO Associated objective(s):

Administrative Conduct Manuals - Licensed Operators/STA Performance Terminal Describe the general requirements for controlling maintenance and modification activities, including:

a. Whose signature is required in order to begin maintenance.

b. What to do if special requirements are involved such as heavy loads, scaffolds, welding, grinding, fire permits, challenges to containment, or barriers.

ILO 2018 NRC Exam Page: 271 of 282 Question 96

97 K/A Importance: 3.2/4.2 Points: 1.00 S97 Difficulty: 3.00 Level of Knowledge: High Source: BANK 73007 The plant is operating at 75% power with B RRMG Scoop Tube LOCKED, with the following recirculation parameters:

Loop A Loop B RRMG set speed 63% 63%

Jet Pump Flow 36 Mlbm/hr 36 Mlbm/hr Total Core Flow 72 Mlbm/hr Following a recirculation flow adjustment, power has lowered to 65% with the following recirculation parameters:

Loop A Loop B RRMG set speed 49% 63%

Jet Pump Flow 27 Mlbm/hr 38 Mlbm/hr Total Core Flow 63 Mlbm/hr Which of the responses below correctly completes the following statement?

These conditions ___(1)___ the Technical Specification 3.4.1, Recirculation Loops Operating LCO, which is based on assumptions which ___(2)___ .

A. (1) MEET (2) prevent excessive component vibration in the reactor vessel B. (1) DO NOT MEET (2) preserve post accident core flow coastdown characteristics C. (1) DO NOT MEET (2) prevent excessive component vibration in the reactor vessel D. (1) MEET (2) preserve post accident core flow coastdown characteristics Answer: B ILO 2018 NRC Exam Page: 272 of 282 Question 97

Answer Explanation:

Tech Specs 3.4.1 Require:

SR 3.4.1.1 States:

Verify recirculation loop jet pump flow mismatch with both recirculation loops in operation is:

a. < 10% of rated core flow when operating at < 70% of rated core flow; and

b. < 5% of rated core flow when operating at > 70% of rated core flow.

Following the adjustment core flow is <70% there fore the < 10% limit applies. Mismatch is > 10%,

therefore LCO is NOT MET.

Therefore the conditions do not meet the LCO. The bases state its for post-accident core flow coast down.

The internal vibrations is a concern when operating with one loop and the other pump speed is greater than 75%.

100% of rated core flow is 100x106lb/hr. Therefore flows in the stem are:

Loop A Loop B Jet Pump Flow 36 Mlbm/hr 36 Mlbm/hr 36% of rated Core Flow 36% of rated Core Flow Jet Pump Flow 27 Mlbm/hr 38 Mlbm/hr 27% of rated Core Flow 38% of rated Core Flow 10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of T.S and how to apply actions. The answer is related immediate actions.

However, to respond the examinee must determine operability first. The answer is not related to TS Safety Limits. The answer is relevant to the entry conditions but is not leading to the answer. While information found above the double line is needed, the correct understanding of the bases for the condition is required. The answer to this questions is based on assessing plant conditions (recirc flow) and evaluating conditions and then applying information from the T.S Bases.

Distractor Explanation:

Distractors are incorrect and plausible based on Answer Explanation and T.S Basis.

Distractor "prevent excessive component vibration in the reactor vessel" is a basis for another recirculation design requirement.

Reference Information:

TS & TSB 3.4.1 ILO 2018 NRC Exam Page: 273 of 282 Question 97

NUREG 1123 KA Catalog Rev. 2 202002 Recirculation Flow Control System G2.2.25 Knowledge of bases in Technical Specifications for limiting conditions for operations and safety limits.

Technical Specifications 3.4.1 Recirculation Loops Operating 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

NRC Question Use (ILO 2018)

BANK High SRO Associated objective(s):

Reactor Recirculation System Cognitive Terminal Analyze a component status change and describe its effect on Reactor Recirculation system operation. (understanding).

ILO 2018 NRC Exam Page: 274 of 282 Question 97

THIS PAGE LEFT BLANK ILO 2018 NRC Exam Page: 275 of 282 Question 98

98 K/A Importance: 2.0/3.8 Points: 1.00 S98V2 Difficulty: 3.00 Level of Knowledge: Low Source: BANK 75026 A discharge permit has been submitted to the Shift Manager for approval. The permit indicates the water to be discharged is potentially contaminated and will be discharged from an area subject to radiological restrictions. The actions required by MRP18, Release of Potentially Clean Fluids, have been completed.

Upon deciding to approve the discharge permit, the Shift Manager should assign an expiration not to exceed:

A. 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> B. 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> C. 5 days D. 30 days Answer: B ILO 2018 NRC Exam Page: 276 of 282 Question 98

Answer Explanation:

Per MCE 06 NON-RADIOLOGICAL ENVIRONMENTAL PROTECTION, pumping of sources within an area subject to radiological restrictions should be assigned an expiration not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from the time the permit is approved.

10 CFR 55.43(b)(4) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because this question is about radiation values used for a SRO responsibility to determine a time limit that is NOT a share responsibility with an RO or are values that are required to be known by the RO for radiological safety stay times.

Distractor Explanation:

A/C/D. Incorrect because of answer explanation and plausible because times listed are acceptable for other conditions in MCE 06.

Reference Information:

MCE06, 5.1.13 Plant Procedures MCE06 NUREG 1123 KA Catalog Rev. 2 G2.3.6 Ability to approve release permits 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

NRC Exam Usage ILO 2012 Exam NRC Question Use (ILO 2018)

BANK Low SRO Associated objective(s):

Senior Reactor Operator Qualification Card Performance Terminal Approve an effluent pump permit including determination of permit duration ILO 2018 NRC Exam Page: 277 of 282 Question 98

99 K/A Importance: 2.4/4.4 Points: 1.00 S99 Difficulty: 3.00 Level of Knowledge: Low Source: BANK 72987 An ALERT Emergency Action Level has been declared. Which facilities MUST be activated?

A. TSC ONLY.

B. TSC and OSC ONLY.

C. TSC, OSC, and EOF ONLY.

D. TSC, OSC, EOF, and JIC.

Answer: C ILO 2018 NRC Exam Page: 278 of 282 Question 99

Answer Explanation:

Per EP-103, TSC, OSC, and EOF must be activated on step 4 of the Alert - Checklist for Immediate Actions.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of Emergency Classification from EP-101, it is not related to immediate actions, and the entry conditions are not relevant or leading to the answer. To answer this question, the examinee must have knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps. In this case EP-103 and applicable ALERT Checklist.

Distractor Explanation:

Distractors are incorrect and plausible because:

A/B. The EOF must be activated for an ALERT per EP-103, and not for other classifications D. The JIC is not activated until a SAE.

Reference Information:

EP-103, page 3, section 4.3 EP-103001, ALERT Checklist Plant Procedures EP-103 NUREG 1123 KA Catalog Rev. 2 G2.4.38 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage ILO 2012 Exam NRC Question Use (ILO 2018)

BANK Low SRO Associated objective(s):

Emergency Procedures/ Plan Cycle 12-1 Objectives Cognitive Enabler Explain the actions for an Alert SRO Objectives Cognitive Terminal Given a copy of EP-103, discuss SRO actions for an Alert ILO 2018 NRC Exam Page: 279 of 282 Question 99

100 K/A Importance: 4.2/4.2 Points: 1.00 S100V2 Difficulty: 2.00 Level of Knowledge: High Source: MODIFIED 75926 While operating at 100% power.

• 3D8 DIV I/II OFF GAS RADN MONITOR UPSCALE alarms.

• Off gas flow rate is 105 SCFM.

And Which of following is the CRS REQUIRED to direct?

A. Preform Rapid Power Reduction B. Reduce Reactor Power using the GOP.

C. Reset the affected Radiation Monitor(s) in the Relay Room to clear 3D8.

D. Place the Mode Switch in Shutdown AND Close or Verify Closed MSIVs and main steam line drains.

Answer: B ILO 2018 NRC Exam Page: 280 of 282 Question 100

Answer Explanation:

The calculated release rate using provided conversion sheet is 278 mCi/sec / 251 mCi/sec and RISING.

Based on this indication levels are CURRENTLY bellow TS limits. Entry into the AOP 20.000.07 Fuel Cladding Failure is required by ARP 3D8.

The candidate must diagnose the trend and recognize that there is a small fuel cladding failure.

The only Actions statement that is met for AOP 20.000.07 is to reduce power using 22.000.03 (Condition B) 10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of AOP actions for a small fuel cladding failure, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this questions is based on assessing plant conditions and then applying the requirements of the AOP.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The ARP requires that affected Radiation Monitor(s) must be reset in the Relay Room to clear 3D8 only if less than the setpoint.

C. This is the Immediate action for a gross fuel failure. This is not a gross failure as per NOTE 1 on page 4 of 20.000.07 D. This is the Immediate action for a gross fuel failure with a few parts of conditions A. This is not a gross failure as per NOTE 1 on page 4 of 20.000.07 Reference Information:

AOP 20.000.07 pg 5 ARP 3D8 Plant Procedures 03D008 20.000.07 NUREG 1123 KA Catalog Rev. 2 G2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material Technical Specifications 3.7.5 Main Condenser Offgas 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage ILO 2012 Exam NRC Question Use (ILO 2018)

High MODIFIED SRO Associated objective(s):

Off Gas Performance Terminal Analyze Off Gas response using control room indications and annunciators.

ILO 2018 NRC Exam Page: 281 of 282 Question 100

ILO 2018 NRC Exam Page: 282 of 282 Question 100

AUTO ACTIONS NOTE: The system that had the critical failure will not automatically restart.

1. Failure of any software module designated as critical will cause an automatic failover of the system.

NOTE: A total loss of VAS indication on H11-P603 OR a loss of data acquisition MUX A and C may meet the loss of monitoring criteria for EP-101 EAL SU3.1, or SA3.1.

INITIAL RESPONSE NOTE: Visual Annunciator System (VAS) Multiplexer (MUX) failure may occur that results in a continuous rebooting of the MUX or false alarms in the Control Room. In either case the MUX not working properly needs to be identified promptly. Upon receipt of alarm 3D19, ANNUNCIATOR SYSTEM TROUBLE, prompt identification of the associated MUX source can be performed using VAS screens as described.

1. Upon receipt of , alarm MUX related events will be noted by:

• At least 2 SOERs will be generated VAS HARDWARE SYSTEM TROUBLE (848C97), VAS SOFTWARE SYSTEM TROUBLE (849C97).

• Alarm Screen (TOC=ALARM) will display at the top part (VAS SYSTEM) specific alarm as to which MUX failed.

• Alarm Screen (TOC=ALARM) will display at the bottom part (INPUT HEALTH) 1280 points in alarm (UNHEALTHY).

• Annunciator Status screen (TOC=STATUS) will show the status of the failed MUX (realtime)

• VAS System Software Trouble Overview (TOC=SWTRBL) will show the status for the failed MUX processor (realtime)

• VAS System Hardware Trouble Overview (TOC=HWTRBL) will show the status of the Failed MUX and its associated Watchdog Timer Card status.

If the initiating condition is a MUX failure and the system does not auto recover or fails again and does not recover within 15 minutes and Action A fails to recover the MUX, remove the failed MUX from service in accordance with 23.621, "Main Control Room Annunciator And Sequence Recorder."

• System running ANNUNCIATOR Primary CPU only Failure of any Critical

• HW Trouble SYSTEM Auto Actions Software Module will SOER in alarm TROUBLE cause an automatic failover

• Hardware System Trouble (HWTRBL)

Review Annunciator

• Software System System Trouble Trouble (SWTRBL) System Running Restart Overview (TROUBLE)

• SOER Screen (SOER) properly with Primary Y Backup only ? CPU N

Initiating Event Analysis H11-P603 Backup TOTAL Y Loss of VAS CPU Y LOSS OF VAS Alarm ? available ?

Contact N Ground Fault Y Maintenance - I&C Detection ?

for ground isolation Initiate N

CARD N

H21-P827 Cabinet Open Y

Temperatures ? Cabinet Doors Close or Verify Closed affected cabinet Circuit Breaker N

Annunciator Verify cabinet System cooling fan is Trouble operating Alarm Clears Hardware System Trouble ?

Y 2 N

Software System Trouble Y 4 TERMINATE

2 Hardware 3

System Trouble Chassis Watchdog Process Nodes /

Tickle Card Timer Network Reset the chassis tickle Associated Watchdog CPU x status = DOWN latches per 23.621 Y Chassis Power ON ? CPU y status = PRIMARY Y N Wait for the Cycle power on chassis tickle alarmed watchdog Network SWCCRB frequency ime timer card chassis Y alarming ?

N N

Watchdog card alarm clears ? MCR SDS PCFMCR alarming Y

Chassis tickle N Use a different alarm 3D19 MCR SDS workstation clear ? for VAS functions Contact Maintenance - I&C for assistance Y TERMINATE

END

DC Sources Operating B 3.8.4 B 3.8 ELECTRICAL POWER SYSTEMS B 3.8.4 DC Sources Operating BASES BACKGROUND The DC electrical power system provides the AC emergency power system with control power. It also provides both motive and control power to selected safety related equipment. As required by 10 CFR 50, Appendix A, GDC 17 (Ref. 1), the DC electrical power system is designed to have sufficient independence, redundancy, and testability to perform its safety functions, assuming a single failure.

The DC electrical power system also conforms to the recommendations of Regulatory Guide 1.6 (Ref. 2) and IEEE-308 (Ref. 3).

The DC power sources provide both motive and control power to selected safety related equipment, as well as circuit breaker control power for the nonsafety related 480 V loads that are connected to 480 V ESF buses. Two center-tapped 260 VDC batteries are provided for Class 1E loads. They are designated 2PA for Division I and 2PB for Division II. Each 260 VDC battery is divided into two 130 VDC batteries connected in series. Each 130 VDC battery section has a battery charger connected in parallel with their respective battery. Each 260 VDC battery has a spare battery charger that can replace either of the normal 130 VDC connected chargers. Each division's two 130 VDC batteries and their chargers are the source of DC control power for that respective division, including the respective EDG. Each 260 VDC source furnishes power to DC motors necessary for shutdown conditions.

During normal operation, the DC loads are powered from the battery chargers with the batteries floating on the system.

In case of loss of normal power to the battery charger, the DC loads are automatically powered from the batteries.

The DC power distribution system is described in more detail in Bases for LCO 3.8.7, "Distribution Systems Operating,"

and LCO 3.8.8, "Distribution Systems Shutdown."

Each battery has adequate storage capacity to carry the required load continuously for approximately 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> (Ref. 11).

FERMI - UNIT 2 B 3.8.4 1 Revision 68

DC Sources Operating B 3.8.4 BASES ACTIONS (continued) division. The 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Completion Time (Required Action A.1) for restoration of an inoperable battery charger allows time to replace the inoperable charger with an OPERABLE spare battery charger, if available. The four hour limit is reasonable based on the remaining capability of the battery to carry the loads for this period. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> limit for Required Action B.1 is consistent with the allowed time for an inoperable DC Distribution System division. The 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Completion Time is based on Regulatory Guide 1.93 (Ref. 6) and reflects a reasonable time to assess unit status as a function of the inoperable DC electrical power subsystem and, if the DC electrical power subsystem is not restored to OPERABLE status, to prepare to effect an orderly and safe unit shutdown.

C.1 If the station service DC electrical power subsystem cannot be restored to OPERABLE status within the required Completion Time, the unit must be brought to a MODE in which overall plant risk is minimized. To achieve this status, the unit must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Remaining in the Applicability of the LCO is acceptable because the plant risk in MODE 3 is similar to or lower than the risk in MODE 4 (Ref. 8) and because the time spent in MODE 3 to perform the necessary repairs to restore the system to OPERABLE status will be short. However, voluntary entry into MODE 4 may be made as it is also an acceptable low-risk state.

Required Action C.1 is modified by a Note that states that LCO 3.0.4.a is not applicable when entering MODE 3. This Note prohibits the use of LCO 3.0.4.a to enter MODE 3 during startup with the LCO not met. However, there is no restriction on the use of LCO 3.0.4.b, if applicable, because LCO 3.0.4.b requires performance of a risk assessment addressing inoperable systems and components, consideration of the results, determination of the acceptability of entering MODE 3, and establishment of risk management actions, if appropriate. LCO 3.0.4 is not applicable to, and the Note does not preclude, changes in MODES or other specified conditions in the Applicability that are required to comply with ACTIONS or that are part of a shutdown of the unit.

FERMI - UNIT 2 B 3.8.4 4 Revision 68