Text

As-Administered Written Examination and Answer Key
	ML23012A168
Person / Time
Site:	Monticello
Issue date:	10/25/2022
From:	; Nuclear Management Co, Xcel Energy
To:	; NRC/RGN-III/DORS/OB
	Iskierka-Boggs T
Shared Package
ML21188A296	List: ML23012A165; ML23012A166; ML23012A168; ML23012A169;
References
	Download: ML23012A168 (1)
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2022 MONTICELLO ILT NRC EXAM - KEY

1.

The plant was at rated conditions when #11 Recirc Pump tripped.

C.4-B.01.04.A (TRIP OF ONE RECIRC PUMP) is being performed

The OATC closes MO-2-53A (11 Recirc Pump Discharge)

Which is correct concerning the action taken above?

MO-2-53A A. should have been left open to prevent thermal clamping.

B. was closed to prevent reverse flow and backward pump rotation.

C. was closed to prevent APRM setpoints from being lower than desired.

D. should have been left open to ensure APRM setpoints are accurate for total core flow.

CORRECT ANSWER:

B JUSTIFICATION: The discharge valve is closed for 5 minutes following a trip. This ensures that there is no reverse flow through the Recirc loop. Reverse flow could cause the pump impeller to rotate backwards. Thus, by closing the valve the pump is allowed to coast down and stop. Also, if enough reverse flow exists, the flow elements will sense this flow and could cause APRM setpoints to be higher than desired.

A is incorrect: The valve is reopened after 5 minutes to prevent this.

C is incorrect: Reverse flow could cause APRM setpoints to be higher that desired.

D is incorrect: When the valve is reopened, the APRM setpoints can make abrupt changes. However, the valve is initially closed.

REFERENCE:

C.4-B.01.04.A 10 CFR 55.41b(5)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

1 GROUP:

1 CATEGORY:

295001 Partial or Complete Loss of Forced Core Flow Circulation K/A:

K3.07 IMPORTANCE:

RO 3.3 COG LEVEL:

1B K/A DESCRIPTION:

Knowledge of the reasons for the following responses or actions as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Recirculation pump discharge/suction valve manipulation.

DIFFICULTY 2

LESSON PL:

MT-ILT-AOP-002L OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

2.

The plant was at rated conditions when a Station Blackout (SBO) occurred. Complete the following statement:

Unnecessary DC loads are disconnected during performance of the SBO procedure to ensure battery capacity is available for ________ of RCIC operation.

A. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B. 1.5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months C. 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months D. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months CORRECT ANSWER:

D JUSTIFICATION: Opening the D111-21 disconnect will remove unnecessary DC load during the SBO event and is required per site analysis in CA-02-179 to ensure battery capacity is available for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months of RCIC operation.

A is incorrect: Plausible time requirement associated with 60 minute capacity of the #17 Battery.

B is incorrect: Plausible time requirement associated with 30 minute load shedding and 60 minute capacity of the #17 Battery.

C is incorrect: Plausible time requirement for HPCI operation during a SBO. During a SBO the HPCI suction is transferred to the CSTs at the 3 hour3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months mark to ensure suction pipe temperature of 170°F is not exceeded.

REFERENCE:

C.4-B.09.02.A 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

1 GROUP:

1 CATEGORY:

295003 Partial or Complete Loss of AC Power K/A:

AK2.07 IMPORTANCE:

RO 3.7 COG LEVEL:

1 B K/A DESCRIPTION:

Knowledge of the relationship between PARTIAL OR COMPLETE LOSS OF AC POWER and the following systems or components: DC electrical distribution system.

DIFFICULTY 3

LESSON PL:

MT-ILT-AOP-024L OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

3.

The plant is shutdown for a planned maintenance outage with "A" RHR in shutdown cooling.

The Control Room Operator has identified the following changes in plant conditions:

SRM Channels 23 and 24 indications are trending down

Service Water Effluent monitor indication is trending down

RBCCW radiation monitor indication is trending down Given the above conditions, which Abnormal Procedure should be entered AND why?

A. LOSS OF A 125 VDC BUS for a loss of D-11 (Div 1 125 VDC Distribution Panel)

B. LOSS OF A 125 VDC BUS for a loss of D-21 (Div 2 125 VDC Distribution Panel)

C. LOSS OF A 24 VDC BUS for a loss of D-15 (Div 1 24 VDC Distribution Panel)

D. LOSS OF A 24 VDC BUS for a loss of D-25 (Div 2 24 VDC Distribution Panel)

CORRECT ANSWER:

D JUSTIFICATION: The loads listed are supplied from D-25. This is MNGP specific OE. A loss of a 24 VDC Battery charger went unnoticed for a period of time resulting in the above indications A is incorrect: Plausible power supply but wrong division and voltage level for the equipment listed.

B is incorrect: Plausible power supply but wrong voltage level for the equipment listed.

C is incorrect: Plausible 24 VDC power supply but wrong division for the equipment listed.

REFERENCE:

B.09.11-05 10 CFR 55.41b(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

1 GROUP:

1 CATEGORY:

295004 Partial or total loss of DC K/A:

2.2.44 IMPORTANCE:

RO 4.2 COG LEVEL:

1 F K/A DESCRIPTION:

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

DIFFICULTY 3

LESSON PL:

M8107L-041 OBJECTIVE:

5

2022 MONTICELLO ILT NRC EXAM - KEY

4.

The plant was at rated conditions when a main turbine generator trip and reactor scram occurred. Given the CURRENT conditions:

Reactor pressure is 810 psig and slowly rising

Main Turbine speed is 1780 rpm and slowly lowering

C.4-A (REACTOR SCRAM) Part B has NOT yet been performed Based on current plant conditions; what is the expected equipment status below?

MAIN TURBINE BYPASS VALVES MAIN TURBINE AUX OIL PUMP A.

OPEN ON B.

OPEN OFF C.

CLOSED ON D.

CLOSED OFF CORRECT ANSWER:

D JUSTIFICATION: With reactor pressure at 810 psig and below both of the normal pressure regulator setpoints

(~905/915) the bypass valves would both be closed at this time. The Aux oil pump is required for pressure control because the main turbine eventually will not be able to maintain oil pressure. Above 1600 rpm the main turbine can supply sufficient oil pressure so without BOP actions taken the pump would remain off.

A is incorrect: The Bypass valves would not be open and the AOP would be off.

B is incorrect: The Bypass valves would not be open.

C is incorrect: The AOP would be off.

REFERENCE:

C.4-A 2204 10 CFR 55.41(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2010 NRC Exam #4 - Minor edits to stem TIER:

1 GROUP:

1 CATEGORY:

295005 Main Turbine Generator Trip K/A:

AK1.01 IMPORTANCE:

RO 4.3 COG LEVEL:

2RI K/A DESCRIPTION: Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to the MAIN TURBINE GENERATOR TRIP: Reactor pressure control.

DIFFICULTY 3

LESSON PL:

M8107L-048 OBJECTIVE:

9.e

2022 MONTICELLO ILT NRC EXAM - KEY

5.

The plant was at rated conditions when a reactor scram occurred. C.4-A (REACTOR SCRAM) directs the OATC to verify that the Recirc pumps have runback to minimum speed or have tripped.

What is the reason this action is performed?

A. To prevent NPSH limits from being exceeded on the Recirc pump suctions.

B. To eliminate potential for neutron flux oscillations if an ATWS were to occur.

C. To eliminate potential for neutron flux oscillations while the reactor is shutting down.

D. To prevent the recirc pumps from running at shutoff head due to the drop in core d/p.

CORRECT ANSWER:

A JUSTIFICATION: The Feedwater/Recirc flow interlock ensures that Recirc flow remains at <30% if FW flow is <20%.

The FW/Recirc Flow interlock assures adequate sub-cooling of the recirculation pump suction water such that the pump NPSH requirement is satisfied. This interlock is dual function when power is raised or lowered. Following the scram, a reduction of steam and feed flow will occur. When feed flow lowers to <20% the Recirc pumps automatically runback to 30% from the FW/Recirc flow interlock.

B is incorrect: During an ATWS condition the recirc pumps are tripped and would not be allowed to remain at minimum speed.

C is incorrect: Flow and power conditions during a scram should just reduce. Assuming all rods insert, power would not be high enough to be concerned with neutron flux oscillations.

D is incorrect: This is the opposite effect as the pump would be more prone to a runout condition.

REFERENCE:

C.4-A ARP C.6-004-C-3/4 B.05.08.01 10 CFR 55.41b(7, 10)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - Edits to stem and choices TIER:

1 GROUP:

1 CATEGORY:

295006 Scram K/A:

AK3.06 IMPORTANCE:

RO 3.7 COG LEVEL:

1B K/A DESCRIPTION:

Knowledge of the reasons for the following responses or actions as they apply to SCRAM:

Recirculation Pump speed reduction.

DIFFICULTY 2

LESSON PL:

MT-ILT-AOP-001L OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

6.

A significant event has occurred that required the crew to establish plant control at ASDS Control Panel C-292.

Which of the following actions can be performed with B RHR from C-292?

A. Raise RPV water level in the LPCI Mode.

B. Lower Torus water temperature in the Torus Cooling Mode.

C. Lower RPV pressure/temperature in the Shutdown Cooling Mode.

D. Lower Drywell pressure/temperature in the Containment Spray Mode.

CORRECT ANSWER:

B JUSTIFICATION: To limit Torus water temperature and Primary Containment pressure, the ASDS uses the Loop 12 RHR System in the Torus Cooling mode to remove Reactor decay heat from the Torus. In the Torus Cooling mode, suction is taken from the Torus and 12 RHR Pump circulates flow through the 12 RHR Heat Exchanger and back to the Torus. The 12 RHR Service Water System is used to remove the decay heat through the RHR Heat Exchanger.

A is incorrect: RPV water level can only be raised using Core Spray from C-292.

C is incorrect: Although a feed and bleed could be performed using SRVs, the Shutdown Cooling Mode cannot be performed from C-292.

D is incorrect: RHR B cannot perform Containment Sprays from C-292.

REFERENCE:

B.5.17-02 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - Edits to stem and choices TIER:

1 GROUP:

1 CATEGORY:

295016 Control Room Abandonment K/A:

AA2.04 IMPORTANCE:

RO 4.0 COG LEVEL:

1F K/A DESCRIPTION:

Ability to determine or interpret the following as they apply to Control Room Abandonment:

Suppression Pool Temperature.

DIFFICULTY 3

LESSON PL:

M8107L-083 OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

7.

Complete the following statement concerning system component temperatures on a loss of RBCCW from rated power.

The (1) system must be shutdown prior to the outlet temperature of the (2) exceeding 120° F to protect downstream piping from over temperature.

(1)

(2)

A.

CRDH CRD Pump Gear Box oil Cooler B.

FPCC Fuel Pool Cooling Heat Exchanger C.

Recirc Reactor Recirc Pump Seal Coolers D.

RWCU Non-regenerative Heat Exchanger CORRECT ANSWER:

B JUSTIFICATION: The FPCC heat exchanger outlet temp must be monitored and FPCC system removed from service before the heat exchanger outlet temp reaches 120 F. The FPCC Heat Exchanger outlet piping is not analyzed for temps above 120°F.

A is incorrect: Plausible RBCCW cooled component; however, this system doesnt have temperature concerns for downstream piping.

C is incorrect: Plausible RBCCW cooled component, however, this system doesnt have temperature concerns for downstream piping.

D is incorrect: Plausible RBCCW cooled component; however, The temperature limit for RWCU is 140°F at the NRHX.

REFERENCE:

C.4-B.02.05.A 10 CFR 55.41b(10)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2020 NRC Exam #7 - Previous two NRC Exams TIER:

1 GROUP:

1 CATEGORY:

295018 Partial or Complete Loss of CCW K/A:

AA1.07 IMPORTANCE:

RO 3.6 COG LEVEL:

1 B K/A DESCRIPTION: Ability to operate or monitor the following as they apply to Partial or Complete Loss of CCW: Fuel pool cooling and cleanup system.

DIFFICULTY 3

LESSON PL:

MT-ILT-AOP-004L OBJECTIVE:

6

2022 MONTICELLO ILT NRC EXAM - KEY

8.

The plant is at rated conditions when a Service Air leak occurs. Given the following:

C.4-B.08.04.01.A (LOSS OF INSTRUMENT AIR) has been entered

Instrument Air (IA) header pressure is 81 psig and lowering

The BOP places the handswitch for CV-1474 (SERV AIR ISOL CV) in CLOSE Which is a correct reason for the BOP placing the handswitch for CV-1474 in CLOSE?

A. CV-1474 should have automatically closed at 90 psig.

B. This is an IMMEDIATE ACTION of C.4-B.08.04.01.A.

C. CV-1474 does NOT have an AUTOMATIC close feature.

D. This prevents CV-1474 from re-opening if air pressure begins to rise.

CORRECT ANSWER:

D JUSTIFICATION: If air header pressure is decreasing, CV-1474 is closed to isolate Service Air which could be one of the potential sources of air leakage. The valve automatically isolates at 82 psig, but would reopen if header pressure increases above approximately 95-97 psig. Manually closing CV-1474 will prevent cycling of the valve.

A is incorrect: The valve does auto close but not until 82 psig.

B is incorrect: This step is NOT and immediate action.

C is incorrect: It does close automatically, but the operator still is required to close this valve manually.

REFERENCE:

C.4-B.08.04.01.A 10 CFR 55.41b(7,10)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

1 GROUP:

1 CATEGORY:

295019 Partial or Complete Los of IA K/A:

AK2.14 IMPORTANCE:

RO 3.5 COG LEVEL:

2 RI/DR K/A DESCRIPTION:

Knowledge of the relationship between the PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following systems or components: Plant Air system DIFFICULTY 2

LESSON PL:

MT-ILT-AOP-017L OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

9.

The plant was at rated conditions when a transient occurred resulting in a reactor scram.

Actions are being taken to place the plant in Cold Shutdown. Given the following:

A RHR is in Shutdown Cooling

RPV pressure is 70 psig and slowly lowering A LOCKOUT on Bus 15 then occurs and 30 minutes later decay heat has caused RPV pressure to rise to 80 psig.

With the conditions above, which of the following is currently available AND can be used to continue the plant cooldown?

A. RCIC.

B. HPCI.

C. B Shutdown Cooling.

D. Main Turbine Bypass Valves.

CORRECT ANSWER:

A JUSTIFICATION: The candidate must recognize the loss of A Shutdown Cooling due to the bus 15 Lockout and since RPV pressure is now above 75 psig (Shutdown Cooling Interlock) restoring A or B Shutdown Cooling is not possible.

IAW EOP Supplemental procedure C.5-3302 (Alternate Pressure Control) RCIC may be used for alternate pressure control as directed by the CRS in the Pressure Leg of EOP-1100. With the loss of Bus 15 the Div 1 250 VDC battery charger is lost but the battery is still available to operate RCIC with RPV pressure > 65 psig.

B is incorrect: With RPV pressure below 100 psig HPCI would be isolated and not available.

C is incorrect: Since RPV pressure is now above 75 psig (Shutdown Cooling Interlock) restoring A or B Shutdown Cooling is not possible.

D is incorrect: With the Bus 15 LOCKOUT, TBVS would be unavailable due to a loss of the Aux Oil Pump.

REFERENCE:

B.03.04-02 C.5-3302 10 CFR 55.41b(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2015 NRC Exam #29 - Minor edits to stem TIER:

1 GROUP:

1 CATEGORY:

295021 Loss of Shutdown Cooling K/A:

AA2.06 IMPORTANCE:

RO 4.2 COG LEVEL:

2 RI K/A DESCRIPTION: Ability to determine or interpret the following as they apply to LOSS OF SHUTDOWN COOLING:

Reactor pressure.

DIFFICULTY 3

LESSON PL:

MT-ILT-AOP-007L OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

10.

The plant is in a Refuel Outage with fuel being loaded into the core.

Which of the following would be considered a refueling error?

A. The lug on the fuel bundle bail points AWAY from the control rod.

B. The base of the fuel bundle serial number is NEAREST the control rod.

C. The channel fastener cap screw is pointed at the CENTER of the control cell.

D. The grapple is LATCHED to an out of sequence in-core fuel bundle (No upward motion has occurred).

CORRECT ANSWER:

A JUSTIFICATION: The lug should point towards the center of the control cell, not away.

B is incorrect: This is proper orientation.

C is incorrect: This is proper orientation.

D is incorrect: Latching an out of sequence fuel bundle is not a refuel error as long as it is not moved (raised). A component is considered to be picked up when the grapple is latched and starts upwards motion.

REFERENCE:

D.2-05 10 CFR 55.41b(13)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

1 GROUP:

1 CATEGORY:

295023 Refueling Accidents K/A:

AK1.04 IMPORTANCE:

RO 3.4 COG LEVEL:

1 F K/A DESCRIPTION:

Knowledge of the operational implications and/or cause effect relationships of the following concepts as they apply to REFUELING ACCIDENTS: Fuel positioning DIFFICULTY 3

LESSON PL:

M8107L-019 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

11.

The plant was operating in MODE 2 with a startup in progress. A large leak develops in the Reactor Vessel Bottom Head and the following timeline of events occur:

TIME Reactor Level Reactor Pressure Drywell pressure 00:00:00 30 inches 900 psig 0.5 psig 00:05:00 9 inches 880 psig 0.7 psig 00:10:00 -15 inches 840 psig 1.3 psig 00:15:00 -30 inches 800 psig 1.9 psig 00:20:00 -40 inches 760 psig 2.1 psig 00:25:00 -47 inches 720 psig 2.3 psig Based on the above timeline, which of the following is correct?

A. ONLY a Partial Group 2 Isolation occurred at 00:05:00.

B. All Main Steam Isolation Valves are closed at 00:10:00.

C. Both Emergency Diesel Generators are running at 00:15:00.

D. Division 1 RHR was selected for LPCI Loop injection path at 00:25:00.

CORRECT ANSWER:

C JUSTIFICATION: The EDGs get an automatic start signal at 1.84 psig in the drywell and will be running at time 00:15:00 A is incorrect: A Full Group 2 Isolation would occur at this time.

B is incorrect: With the plant in Mode 2 the Mode Switch would be in STARTUP. Therefore the <840 psig in RUN Group 1 isolation will be bypassed.

D is incorrect: Plausible as LPCI loop selection takes place at -47 and 1.84 psig, and previous setpoint was 2.0 psig. Division 2 is the default loop for LPCI injection and since the leak is in the bottom head, Division 1 would NOT be selected.

REFERENCE:

B.09.08.01 10 CFR 55.41b(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2015 NRC Exam #11 TIER:

1 GROUP:

1 CATEGORY:

295024 High Drywell Pressure K/A:

EK2.06 IMPORTANCE:

RO 4.2 COG LEVEL:

2RI K/A DESCRIPTION:

Knowledge of the relationship between HIGH DRYWELL PRESSURE and the following systems or components: Emergency generators.

DIFFICULTY 2

LESSON PL:

M-8107L-042 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

12.

The plant is at rated conditions when a Group 1 isolation results in a scram and numerous control rods FAIL to insert. C.5-2007 (FAILURE TO SCRAM) has been entered.

The OATC chooses to perform Part D (RESCRAM CONTROL RODS) of procedure C.5-3101 (ALTERNATE ROD INSERTION).

Which of the following is CORRECT for the performance of Part D?

A. Low-Low Set will be disabled during a portion of Part D.

B. RPV water level must be > -47 inches to perform Part D.

C. BOTH divisions of RPS power are required to perform Part D.

D. CRDH flow will add additional positive reactivity once the scram is reset.

CORRECT ANSWER:

A JUSTIFICATION: In order for Lo-Lo Set to operate automatically, a scram signal must be present. C.5-3101 CAUTION: Low-Low Set will not operate with the scram reset.

B is incorrect: The ATWS signal that would be present <-47 inches can be bypassed by opening 2 breakers using Part D.

C is incorrect: Only one division of RPS would be required to reset the scram and perform Part D.

D is incorrect: When the scram is reset CRD injects less water into the core.

REFERENCE:

C.5-3101 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2009 NRC Exam #17 TIER:

1 GROUP:

1 CATEGORY:

295025 High Reactor Pressure K/A:

2.1.32 IMPORTANCE:

RO 3.8 COG LEVEL:

1P K/A DESCRIPTION:

Ability to explain and apply system precautions, limitations, notes or cautions.

DIFFICULTY 3

LESSON PL:

MT-ILT-EOP-007L OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

13.

The plant was at rated conditions when a severe ATWS condition occurred. The following indications were noted at 0800:

Of the times below, when will the Torus Heat Capacity (Detail M) be exceeded FIRST?

(Assume Torus temperature rate of change, Torus water level and Reactor pressure remain constant. Detail M is on the following page).

A. 0845 B. 0840 C. 0836 D. 0829 CORRECT ANSWER:

B JUSTIFICATION: At 0840 Torus temperature will have risen to (40 min x 2.9F/min + 88.1F = 204.1F) With Rx Pressure at approximately 1000 psig and using the solid (-4 to +3) Torus level curve, Detail M will be exceeded.

A is incorrect: After 45 minutes Torus water temperature will be 218.6F. This exceeds Detail M but wouldnt be first.

C is incorrect: After 36 minutes Torus water temperature will be 192.5F. This would exceed Detail M only if using the incorrect middle dotted line.

D is incorrect: After 29 minutes Torus water temperature will be 172.2F. This would be the first to exceed Detail M only if using the incorrect lower dotted Torus Water level curve.

REFERENCE:

C-5-1200 Detail M 10 CFR 55.41b(7)

REFERENCE PROVIDED DURING EXAM:

Embedded - Figure M QUESTION SOURCE:

New TIER:

1 GROUP:

1 CATEGORY:

295026 Supp Pool High Water Temp K/A:

EK1.04 IMPORTANCE:

RO 3.5 COG LEVEL:

3 SPK K/A DESCRIPTION: Knowledge of the operational implications and/or cause effect relationships of the following concepts as they apply to the SUPP POOL HIGH WATER TEMP: Suppression pool level DIFFICULTY 3

LESSON PL:

MT-ILT-EOP-003L OBJECTIVE:

7.b

2022 MONTICELLO ILT NRC EXAM - KEY

14.

The plant was at rated conditions when an event occurred in the Drywell (DW) resulting in the following conditions:

DW pressure is 12.2 psig and rising slowly

DW temperature is 225°F and rising slowly

Torus water level is +1 inch and rising slowly

Torus water temperature is 95°F and rising slowly

4 RHR pumps are running and A RHR is in the Torus Spray mode At this point, the CRS directs that B RHR be placed in the Drywell Spray mode. Which of the following statements describes why this action is needed?

Drywell sprays A. ARE initiated above 12 psig DW pressure to restore and maintain pressure suppression capability.

B. CAN NOT be initiated below 12 psig DW pressure because RHR NPSH limits would be exceeded.

C. CAN NOT be initiated below 12 psig DW pressure because de-inertion of the primary containment could occur.

D. ARE initiated above 12 psig DW pressure to preclude chugging that could cause fatigue failure in the downcomers.

CORRECT ANSWER:

D JUSTIFICATION: Drywell sprays are initiated if drywell pressure exceeds 12 psig to preclude a phenomenon known as chugging.

Chugging is the cyclic condensation of steam at the downcomer openings of the drywell vents. When a steam bubble collapses at the exit of the downcomers, the rush of water drawn in the downcomers to fill the void induces stress at the junction of the downcomers and the vent header. Repeated application of such stresses could cause fatigue failure of these joints.

A is incorrect: Pressure suppression capability would not be a concern until DW pressure exceeds 25 psig. Although one of the goals of spraying the DW is to prevent losing this capability, its not the reason for doing it at 12 psig and the restore in this distracter implies that it is currently being exceeded which it isnt.

B is incorrect: RHR NPSH limits are not being exceeded as torus water temperature would need to be in excess of 200°F for this to be true.

C is incorrect: The DW Spray limit is only a concern at much lower DW temperatures.

REFERENCE:

C.5.1-1001 10 CFR 55.41(5)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2020 NRC Exam #11 - Previous 2 NRC Exams TIER:

1 GROUP:

1 CATEGORY:

295028 High Drywell Temperature K/A:

EK3.03 IMPORTANCE:

RO 3.8 COG LEVEL:

2DR K/A DESCRIPTION:

Knowledge of the reasons for the following responses or actions as they apply to HIGH DRYWELL TEMPERATURE: Drywell spray.

DIFFICULTY 3

LESSON PL:

MT-ILT-EOP-003L OBJECTIVE:

7.b

2022 MONTICELLO ILT NRC EXAM - KEY

15.

The plant was at rated conditions when a major event occurred including a Torus leak.

Torus water level is -1 foot and lowering

RCIC is injecting to the RPV at rated flow Which Torus level below would you FIRST report that the RCIC exhaust is discharging into the Torus air space?

A. - 3.5 feet B. - 4.0 feet C. - 4.5 feet D. - 6.0 feet CORRECT ANSWER:

B JUSTIFICATION: The RCIC turbine exhaust discharges to the torus water space at -3.7 ft.

A is incorrect: Plausible level as the downcomers become uncovered at -3.3 feet.

C is incorrect: Plausible number as required actions must be taken when Torus level is +4.2 feet.

D is incorrect: Plausible level as SRVs become uncovered at - 5.9 feet.

REFERENCE:

C.5.1-1001 10 CFR 55.41b(8)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

1 GROUP:

1 CATEGORY:

295030 Low Suppression Pool Level K/A:

EA1.02 IMPORTANCE:

RO 3.8 COG LEVEL:

1 F K/A DESCRIPTION:

Ability to operate or monitor the following as they apply to Low Suppression Pool Level: RCIC.

DIFFICULTY 2

LESSON PL:

MT-ILT-EOP-001L OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

16.

The plant was at normal rated conditions when a LONOP and LOCA occurred.

RBCCW has been restored

CRD-14 (CRD CHARGING WATER TO ACCUMS) is OPEN

C.5-3204 (RPV MAKEUP WITH CRD) is required for RPV level control Which is correct concerning RPV level if CRD pumps are restarted with the conditions above?

A. A pressure transient could cause a LOWER indicated RPV water level.

B. A pressure transient could cause a HIGHER indicated RPV water level.

C. RPV level would not be affected; CRD pumps will NOT start following a LONOP.

D. RPV level would not be affected; CRD-141 (RPV BACKFILL ISOL) auto CLOSES on a scram.

CORRECT ANSWER:

A JUSTIFICATION: Starting a CRD pump with CRD-14 left open and CRD-141 open has caused pressure transients in the Reference Leg Backfill System, resulting in a low indicated RPV water level. This is MNGP specific OE.

B is incorrect: The pressure transient could cause a lower indicated water level.

C is incorrect: Following a LONOP, 15 and 16 Bus will re-energize from 1AR or the EDGs. These pumps can be restarted once RBCCW is restored.

D is incorrect: Under normal conditions, this valve is open and does NOT auto close on a scram.

REFERENCE:

C.5-3204 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

1 GROUP:

1 CATEGORY:

295031 RPV low water level K/A:

EA1.10 IMPORTANCE:

RO 3.6 COG LEVEL:

2 DR K/A DESCRIPTION:

Ability to operate and/or monitor the following as they apply to RPV LOW WATER LEVEL:

Control rod drive hydraulic system.

DIFFICULTY 3

LESSON PL:

MT-ILT-EOP-002L OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

17.

The plant was at rated conditions when a scram and ATWS condition occurred. The following conditions are present:

Reactor power 20% and lowering

The Main Condenser is unavailable

RPV pressure is cycling on Lo-Lo Set

Torus water temperature is 95°F and rising

RPV water level is being maintained -150 to -100

Standby Liquid Control (SBLC) System 1 has been initiated The CRS now directs the OATC to report when SBLC tank level reaches 985 gallons.

Which of the following describes the reason for reporting this tank level?

A. This will allow a normal RPV depressurization to begin.

B. This will allow RPV water level to be restored to +9 to +48.

C. This will allow SBLC injection from System 1 to be secured.

D. This will allow the crew to exit C.5-2007 (FAILURE TO SCRAM).

CORRECT ANSWER:

B JUSTIFICATION: ROs are required to report when SBLC tank level lowers to 985 gallons and hot shutdown boron weight has been injected. At this point RPV water level may be restored to +9 to +48 inches.

A is incorrect: In an ATWS, depressurization wouldnt be allowed until Cold Shutdown Boron is injected which is when tank level lowers to 475 gallons.

C is incorrect: C.5-2007 doesnt allow for SBLC injection to be secured until all control rods are inserted.

D is incorrect: C.5-2007 cannot be exited until all control rods are inserted to at least position 04 or if it is determined that the RX will stay shutdown under all conditions.

REFERENCE:

C.5.1-1001 10 CFR 55.41(6)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2010 NRC Exam #17 - Tank level edit in stem due to new EOPs TIER:

1 GROUP:

1 CATEGORY:

295037 Scram Condition Present and Reactor Power Above APRM Downscale or Unknown K/A:

EA2.03 IMPORTANCE:

RO 4.3 COG LEVEL:

2 DR K/A DESCRIPTION: Ability to determine or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: SBLC Tank level.

DIFFICULTY:

3 LESSON PL:

MT-ILT-EOP-007L OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

18.

The plant is at rated conditions. Given the following:

0800: Alarm 4-A-12 (OFF GAS HI RADIATION) is received 0805: A Rapid Power Reduction is performed to 80% reactor power 0815: Main Steam Line Radiation Monitors still indicate above normal radiation levels and are trending up Which action must now be taken for the above conditions and why?

A. Initiate the SBGT system to filter the off-site release.

B. Insert a Reactor scram to minimize the off-site release.

C. Bypass the Off-gas Storage system to isolate the off-site release.

D. Restart an Off-gas Recombiner train to lower off-gas radiation levels.

CORRECT ANSWER:

B JUSTIFICATION: If the high radiation condition is confirmed (as noted by the MSL radiation monitors trending up),

then reduce reactor power per C.4.F. If this action is taken without success, the operator is directed to insert a reactor scram.

A is incorrect: Plausible to think SBGT would filter the release but it doesnt with the release in the steam lines.

C is incorrect: Plausible action as this would be done if off-gas was receiving high hydrogen concentrations but not for high radiation. Also plausible action to take if the Recombiners trip, but that doesnt occur for 30 minutes.

D is incorrect: Plausible action to take but the train trip doesnt occur until 30 minutes have elapsed with a high radiation condition.

REFERENCE:

ARP 4-A-12 10 CFR 55.41b(10)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Significantly Modified - 2015 NRC Exam #23 TIER:

1 GROUP:

1 CATEGORY:

295038 High Off-site Release Rate K/A:

EK3.05 IMPORTANCE:

RO 4.1 COG LEVEL:

1 P K/A DESCRIPTION:

Knowledge of the reasons for the following responses as they apply to HIGH OFF-SITE RELEASE RATE: Reactor Shutdown/Scram DIFFICULTY 2

LESSON PL:

M8107L-009 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

19.

The plant is at rated conditions.

A small lube oil fire is reported in the Recirc MG Set Room

C.4-B.08.05.A (PLANT FIRE) has been entered

You are the Fire Brigade Leader Which is the most effective method you should use to extinguish this fire?

A. CO2 B. Halon C. Pressurized water D. Dry Chemical / Foam CORRECT ANSWER:

D JUSTIFICATION: The examinee must recognize that a lube oil fire is a Class B fire. Class B fires are most effectively extinguished through the use of a dry chemical / Foam extinguisher. At MNGP, the BOP operator is designated as the Fire Brigade Leader.

A is incorrect: CO2 is most effective on Class C electrical fires.

B is incorrect: Halon is most effective on Class C electrical fires.

C is incorrect: Pressurized water is most effective on Class A or combustible fires.

REFERENCE:

A.3-003 A.3-13-A 10 CFR 55.41b(10)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

1 GROUP:

1 CATEGORY:

600000 Plant fire on site K/A:

AK1.02 IMPORTANCE:

RO 3.4 COG LEVEL:

1 F K/A DESCRIPTION:

Knowledge of the operational implications and/or cause and effect relationships of the following as they apply to PLANT FIRE ON SITE: Firefighting methods for each type of fire.

DIFFICULTY 3

LESSON PL:

M8108L-033 OBJECTIVE:

1

2022 MONTICELLO ILT NRC EXAM - KEY

20.

The plant was at rated conditions when grid disturbance resulted in a Main Generator transformer lockout. A fire has erupted in the main generator transformer and the Fire Brigade has been dispatched.

Where can the Main Generator Transformer deluge be initiated from?

A. A control switch on C-300 (Zonalert Control Room Panel).

B. A deluge valve in the Northwest corner of the12 EDG Room.

C. A deluge valve outside the South wall of the Recirc MG Set Room.

D. A deluge valve in the Southeast corner of the East Electrical Equipment Room.

CORRECT ANSWER:

B JUSTIFICATION: Initiation for this transformer is manual and may be accomplished from C-20 (Control Room Fire Panel), or locally at the deluge valve in the 12 EDG Room or from a control switch on the outside of the 12 EDG Building. Locally, this would be directed by the BOP RO who is the designated Fire Brigade Leader.

A is incorrect: This is a Control Room fire alarm panel only.

C is incorrect: The deluge valves in this area are only for the 13 Diesel Generator Room.

D is incorrect: This deluge valve is specifically only for the Recirc MG Set room.

REFERENCE:

B.08.05-01 10 CFR 55.41b(8)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Significantly Modified - 2013 NRC Exam #67 TIER:

1 GROUP:

1 CATEGORY:

700000 GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES K/A:

2.4.26 IMPORTANCE:

RO 3.1 COG LEVEL:

1 S K/A DESCRIPTION:

Knowledge of facility protection requirements, including fire brigade and portable firefighting equipment usage.

DIFFICULTY 3

LESSON PL:

M8107L-010 OBJECTIVE:

6.e

2022 MONTICELLO ILT NRC EXAM - KEY

21.

The plant was in MODE 1 when a Main Feedwater Regulating Valve (MFRV) lockup occurred.

While implementing C.4-B.05.07.A (LOSS OF REACTOR WATER LEVEL CONTROL), the locked MFRV drifts open causing a High RPV Water Level Trip.

1. Which logic initiates this trip?

2. What equipment is protected by this trip?

(1)

(2)

A.

ATWS SRV Bellows B.

ECCS SRV Bellows C.

ATWS Main Turbine blading D.

ECCS Main Turbine blading CORRECT ANSWER:

C JUSTIFICATION: If a MFRV is locked up it may drift open or closed over time. The four ATWS level transmitters drive four master trip units which in turn drive four slave trip units. These slave trip units trip the Main Turbine and Reactor Feed Pumps on high Reactor water level to prevent damage due to excessive carryover. Excessive carryover results in damage to the turbine due to water erosion of turbine blading.

A is incorrect: SRV bellows could experience excessive moisture but this is not the purpose of the trip.

B is incorrect: SRV bellows could experience excessive moisture but this is not the purpose of the trip. Plausible logic choice for high RPV level.

D is incorrect: Plausible logic choice for high RPV level.

REFERENCE:

B.05.06-02 C.4-B.05.07.A Bases 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

1 GROUP:

2 CATEGORY:

295008 High Reactor Water Level K/A:

AK1.01 IMPORTANCE:

RO 3.3 COG LEVEL:

2 RI K/A DESCRIPTION: Knowledge of the operational implications and/or cause and effect relationships of the following concepts as they apply to HIGH REACTOR WATER LEVEL: Moisture carryover.

DIFFICULTY 2

LESSON PL:

M8107L-071 OBJECTIVE:

1

2022 MONTICELLO ILT NRC EXAM - KEY

22.

The plant was at rated conditions when a steam line rupture resulted in the following:

Drywell temperature is 340°F and rising

Attempts to spray the drywell have been unsuccessful

The CRS is directing an RPV Blowdown While performing RPV Blowdown actions, the following indications were noted:

(ALL lights below are extinguished)

Which SINGLE condition below would be a valid reason for the indications shown above?

A. The SRV bellows has RUPTURED.

B. The SRV amber light is BURNT OUT.

C. The SRV is mechanically STUCK CLOSED.

D. The SRV solenoid valve is NOT ENERGIZED.

CORRECT ANSWER:

D JUSTIFICATION: The ADS valve EQ Qualification maximum temperature is 338°F. Since this has been exceeded the ADS valves may fail to open on an ADS initiation or if manually opening the valve. The high drywell temperature and humidity conditions can cause the solenoid valve to short resulting in blow power supply fuses resulting in the solenoid becoming deenergized and the red light remaining off.

A is incorrect: If the bellows ruptures this would only prevent the valve from opening at its safety setpoint.

B is incorrect: With the switch in OPEN the SRV should be open. This normally would cause the red light and amber light to be on. A single failure of the amber light would not cause these indications.

C is incorrect: If the valve was mechanically stuck closed the red light would still be on because the switch is in open.

REFERENCE:

C.5.1-1001 NX-7831-143-1 NX-7831-143-2 10 CFR 55.41(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2010 NRC Exam #14 TIER:

1 GROUP:

2 CATEGORY:

295012 High Drywell Temperature K/A:

AA2.04 IMPORTANCE:

RO 3.8 COG LEVEL:

3 SPR/SPK K/A DESCRIPTION: Ability to determine or interpret the following as they apply to HIGH DRYWELL TEMPERATURE:

System/component operating limitations.

DIFFICULTY 3

LESSON PL:

MT-ILT-EOP-003L OBJECTIVE:

6.b

2022 MONTICELLO ILT NRC EXAM - KEY

23.

A plant startup is in progress and control rod 34-39 is selected to be NOTCHED OUT from Position 10 to Position 12. The RWM Withdraw Limit is Position 16.

After initiating a single ROD OUT NOTCH Sequence, the OATC observes the following:

4-Rod Display indicates Control Rod 34-39 has STOPPED at Position 14

Control Rod Select Matrix Backlighting for Control Rod 34-39 is ON

Full Core Display white select light for Control Rod 34-39 is OFF

The 2 SEC SELECT BLOCK red light alarm is ON Which one of the following CORRECTLY identifies the cause of these indications?

A. The RMCS timer has failed.

B. Rod select power has been lost.

C. The RWM is enforcing a Withdraw Block.

D. The RPIS Reed Switch at Position 11 momentarily closed.

CORRECT ANSWER:

A JUSTIFICATION: An interlock is provided in RMCS to disable rod withdrawal and prevent inadvertent reactivity addition in the event the rod drive timer switch should fail while a rod withdrawal movement is in progress. If such a failure occurred, the interlock would cause the directional control solenoid valves of the selected rod to be disconnected from the drive control busses of the RMCS following a time delay of approximately two seconds. Proper operation of the interlock is demonstrated when the associated rod alarm light located in the full core display is OFF. ARP 5-A-19 (ROD SELECTOR BLOCK TIMER MALFUNCTION) confirmatory indications: Indicating light 2 SEC SELECT BLOCK red light alarm is ON (Panel C-05)

B is incorrect: The backlighting would also be lost.

C is incorrect: The control rod position is still within the limits the RWM.

D is incorrect: This would not cause the white select light to go out.

REFERENCE:

B.05.05-01 ARP 5-A-19 10 CFR 55.41(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2020 NRC Exam #25 - Previous 2 NRC Exams TIER:

1 GROUP:

2 CATEGORY:

295014 Inadvertent Reactivity Addition K/A:

2.4.46 IMPORTANCE:

RO 4.2 COG LEVEL:

3 SPK K/A DESCRIPTION:

Ability to verify that the alarms are consistent with the plant conditions.

DIFFICULTY 3

LESSON PL:

M8107L-032 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

24.

The plant is at rated conditions when a steam leak occurs in the Steam Jet Air Ejector room.

If an automatic Secondary Containment isolation/SBGT initiation occurs, what will happen to the release rate from the RB Exhaust Plenum Room, and why?

The release rate will A. lower because the Turbine Building supply fans will trip.

B. rise because the Plenum Room exhaust fans take suction from the SJAE Room.

C. rise because SJAE Room exhaust will be redirected to the RB Exhaust Plenum Room.

D. lower because the SBGT system will be processing the steam leak from the SJAE Room.

CORRECT ANSWER:

C JUSTIFICATION: V-EF-26 takes suction from the air ejector room and exhausts it to the stack under normal conditions.

This also is the source of stack dilution air. On a secondary containment isolation/SBGT start, the V-EF-26 exhaust is isolated from the stack and re-routed to the RB exhaust plenum room. SBGT discharges to the stack acting as the dilution air. Because the contents of the steam jet air ejector room are exhausted to the plenum room, release rates can be expected to rise.

A is incorrect: The TB fans will trip but release rate will not lower.

B is incorrect: The plenum room exhaust fans will trip on a secondary containment isolation D is incorrect: The release rate will rise.

REFERENCE:

B.08.07-02 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2009 NRC Exam #46 - Edits to stem and choices TIER:

1 GROUP:

2 CATEGORY:

295017 High Offsite Release Rate K/A:

AK3.01 IMPORTANCE:

RO 4.0 COG LEVEL:

1 I K/A DESCRIPTION: Knowledge of the reasons for the following responses or actions as they apply to ABNORMAL OFFSITE RELEASE RATE: System isolations DIFFICULTY 3

LESSON PL:

M8107L-077 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

25.

The plant was at rated conditions when a Reactor scram was inserted due to a high radiation condition in the Reactor Building. Given the following:

Annunciators that are IN alarm:

3-B-55 (REACTOR BLDG EXH PLENUM HI RAD)

259-A-6 (RBV EFFLUENT HIGH RADIATION)

Annunciators that are NOT IN alarm:

3-A-49 (SBGT ANNUNCIATOR)

5-A-1/2 (REAC BLDG VENT & FP RAD CH A/B - HI/LO)

Based ONLY on the alarm conditions above; which is a correct action to take, and why?

A. Restart Secondary Containment ventilation because both SBGT trains failed to start.

B. Depress both SBGT TEST pushbuttons on C-24A/B because both SBGT trains failed to start.

C. Restart Secondary Containment ventilation to ensure a negative pressure exists in the Reactor Building.

D. Depress both SBGT TEST pushbuttons on C-24A/B to ensure a filtered release from the Reactor Building.

CORRECT ANSWER:

D JUSTIFICATION: If a leak is within the Reactor Building and a radioactive release approaches the RBV WRGM Hi-Hi alarm setpoint, then Manually Isolate SCTMT. This procedure isolates SCTMT and auto starts SBGT by pushing the two Test pushbuttons on C-24A/B.

A is incorrect: This action would not be taken because of high radiation in the RB. Additionally the SBGT auto initiation setpoint has not been reached yet as 3-A-49 and 5-A-1/2 are not in alarm.

B is incorrect: The SBGT auto initiation setpoint has not been reached yet as3-A-49 and 5-A-1/2 are not in alarm.

C is incorrect: This action would not be taken because of high radiation in the RB. Additionally, the negative pressure should be established with SBGT.

REFERENCE:

ARP 259-A-6 10 CFR 55.41b(10)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2015 NRC Exam #27 TIER:

1 GROUP:

2 CATEGORY:

295034 Secondary Containment Ventilation High Radiation K/A:

EA1.03 IMPORTANCE:

RO 3.8 COG LEVEL:

3 SPK K/A DESCRIPTION:

Ability to operate or monitor the following as they apply to Secondary Containment Ventilation High Radiation: Secondary containment ventilation.

DIFFICULTY 4

LESSON PL:

MT-ILT-AOP-005L OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

26.

Given the following conditions:

Main Steam Chase temperature is 220°F and rising

Room temperatures in other Reactor Building areas are normal

Numerous Turbine Building rad monitors show a rise in radiation levels Complete the following statements:

PCIS Group __(1)__ Isolation(s) has/have occurred.

Rupture of the Blowout Panels is causing steam to be released to the __(2)__.

(1) (2)

A. IV & V Turbine Operating Floor B. I Turbine Operating Floor C. IV & V Condenser Hotside D. I Condenser Hotside CORRECT ANSWER:

B JUSTIFICATION: The building is designed to withstand an internal pressure of seven inches of water without structure failure and without pressure relief. Analyses have been made for the effect of rupturing lines containing steam, hot water and cool water. Of these, only a failure in the main steam system could release sufficient energy to cause structural failure of the Secondary Containment. The main steam lines are located in the steam tunnel and, consequently, blowout panels are located in the north wall of the steam tunnel wall at elevation 9510 to provide a release path into the Turbine Building. These panels are designed to fall out at a differential pressure of less than 0.25 psi (7 inches of water) and would require resealing if blown. Steam Tunnel High Temperature: High temperature in the vicinity of the main steam lines could indicate a steam line break. To limit the release of coolant and radioactivity, an isolation of Group 1 valves is initiated. The trip setting is 195° to 200°F.

A is incorrect: Group IV and V temperature switches actuate on high temperatures in the torus room, HPCI room and RCIC room ONLY C is incorrect: Group IV and V temperature switches actuate on high temperatures in the torus room, HPCI room and RCIC room ONLY. The blowout panels are located between the Turbine Floor and the Steam Chase. Condenser Hotside is a plausible enclosed location in the turbine building if candidate is unsure about the release directly to the turbine floor.

D is incorrect: The blowout panels are located between the Turbine Floor and the Steam Chase. Condenser Hotside is a plausible enclosed location in the Turbine Building if candidate is unsure about the release directly to the turbine floor.

REFERENCE:

B.04.02-02 10 CFR 55.41b()

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

1 GROUP:

2 CATEGORY:

Secondary Containment High D/P K/A:

EK2.04 IMPORTANCE:

RO 3.5 COG LEVEL:

2 DR K/A DESCRIPTION:

Knowledge of the relationship between SECONDARY CONTAINMENT HIGH D/P and the following systems or components: Blow-out panels.

DIFFICULTY 2

LESSON PL:

M8107L-008 OBJECTIVE:

6

2022 MONTICELLO ILT NRC EXAM - KEY

27.

The plant was operating at rated conditions when a LOCA occurred in the Drywell (DW).

Given the following:

DW pressure is 11 psig and slowly lowering

RPV pressure is 440 psig and slowly lowering

RPV water level is -164 inches and slowly lowering

B RHR System status is shown on the following page Based on the given conditions, complete the following statement:

The Torus Spray/Cooling valves WILL / SHOULD HAVE A. automatically close when DW pressure is < 1.5 psig.

B. automatically close when RPV pressure is < 420 psig.

C. automatically close when RPV water level is < 2/3 core height.

D. automatically closed when RPV pressure went < 460 psig.

CORRECT ANSWER:

C JUSTIFICATION: Based on the given conditions, the examinee must determine that the torus spray valves will automatically close when RPV level goes below -174 (2/3 Core Height). This would be indicative of an instrumentation failure. Keylocked switch (NEUTRAL-MAN OVERRIDE) 10A-S18B will override the 2/3 core height permissive but the picture shows that this hasnt been overridden. These valves are all interlocked to be automatically closed when a LPCI automatic initiation signal is received. Interlocks are designed to prevent diverting LPCI flow to areas other than the Vessel unless it is necessary or unless the LPCI requirements are satisfied. The Torus spray/cooling valves (MO-2006 through MO-2011) cannot be opened when a LPCI initiation signal is present except under the following conditions:

a. Reactor water level is greater than 2/3 core height and

b. Drywell pressure is greater than 1 psig.

A is incorrect: The torus spray valve (MO-2011) will automatically close if DW pressure lowers to <.75 psig.

B is incorrect: This would be true, however the LPCI signal is currently bypassed as indicated by the LPCI Initiation Bypass light being out.

D is incorrect: The LPCI initiation signal pressure used to be 460 psig but now is 420 psig. Plausible to think these valves have already closed on RPV pressure.

REFERENCE:

B.03.04-02 10 CFR 55.41b(7)

REFERENCE PROVIDED DURING EXAM:

Embedded Picture of B RHR on C-03 QUESTION SOURCE:

Bank - 2013 NRC Exam #59 TIER:

2 GROUP:

1 CATEGORY:

203000 RHR/LPCI Injection Mode K/A:

A1.01 IMPORTANCE:

RO 4.4 COG LEVEL:

3 SPR K/A DESCRIPTION: Ability to predict and/or monitor changes in parameters associated with operation of the RHR/LPCI INJECTION MODE including: Reactor water level.

DIFFICULTY 4

LESSON PL:

M8107L-023 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

28.

The plant is in MODE 5 performing a core reload:

The core reload is 50% complete

RPV water level is > 2111 above the top of the RPV flange

Division 1 Shutdown Cooling (SDC) is inoperable and unavailable

12 and 14 RHR pumps are in the SDC mode If 14 RHR pump trips, can fuel loading into the core continue, why or why not?

A. NO, since Division 1 SDC is not available.

B. YES, since one RHR pump is still the SDC mode of operation.

C. YES, if an alternate method of DHR is verified available within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

D. NO, two RHR pumps must be in the SDC mode of operation at all times when moving fuel.

CORRECT ANSWER:

B JUSTIFICATION: One RHR shutdown cooling subsystem shall be operable and in operation in Mode 5 with irradiated fuel in the vessel and water level > 2111 above the flange. Examinee must understand that the critical safety function of DHR is still safe and that fuel movements can still remain a priority With 12 RHR pump still in operation ops can continue to prioritize the reactivity safety function and justify continued fuel movements.

A is incorrect: Division 1 SDC would not be required in this case since 12 RHR pump is running.

C is incorrect: This would be the correct answer if both pumps were lost.

D is incorrect: The required subsystem only needs one pump to be considered operable.

REFERENCE:

TS 3.9.7

< 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS Action 10 CFR 55.41b(10)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - Edits to stem and choices TIER:

2 GROUP:

1 CATEGORY:

205000 Shutdown Cooling System K/A:

2.4.22 IMPORTANCE:

RO 3.6 COG LEVEL:

2 DR K/A DESCRIPTION: Knowledge of the bases for prioritizing safety functions (DHR & Reactivity) during abnormal/emergency operations.

DIFFICULTY 3

LESSON PL:

M8107L-023 OBJECTIVE:

10

2022 MONTICELLO ILT NRC EXAM - KEY

29.

The plant is in MODE 4 with the following conditions:

11 & 13 RHR Pumps are at rated flow in the Shutdown Cooling mode.

12 & 14 RHR Pumps are at rated flow in the Torus Cooling mode.

A small LOCA results in RPV water level lowering to a stable +5 inches.

Which of the following flowrates approximates the TOTAL RHR System flow rate once conditions stabilize?

A. 0 gpm B. 6,000 gpm C. 8,000 gpm D. 16,000 gpm CORRECT ANSWER:

C JUSTIFICATION: A Group 2 PCIS (RPV Level <+9) will close the shutdown cooling isolation valves (MO-2029 & MO-2030. However, this isolation will not affect the torus cooling lineup as a LPCI initiation is not received until -47. With

12 and #14 RHR pumps running, approximate system flow will be 8000 gpm.

A is incorrect: Plausible for complete system isolation.

B is incorrect: Plausible for incorrect rated flow rate of 3000 gpm.

D is incorrect: Plausible for no system isolations.

REFERENCE:

B.03.04-02 10 CFR 55.41b(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

1 CATEGORY:

205000 RHR SDC Mode K/A:

K1.08 IMPORTANCE:

RO 3.7 COG LEVEL:

2 RI K/A DESCRIPTION:

Knowledge of the physical connections and/or cause-effect relationships between RHR SHUTDOWN COOLING MODE and the following: RHR/LPCI DIFFICULTY 3

LESSON PL:

MT-ILT-AOP-007L OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

30.

The plant was at rated conditions when a LOCA occurred. Given the following:

RPV water level lowered to -45" and is now +35" and stable

RPV pressure is 650 PSIG and slowly lowering

Drywell pressure 5 PSIG and rising slowly

Torus level is +1 foot and rising slowly Assuming NO operator actions, which one of the following is correct?

A. The ADS timer is initiated.

B. RCIC suction has auto transferred at a torus level of +2 inches.

C. Core spray pumps are running and injecting to the Reactor vessel.

D. HPCI is injecting at rated flow to the RPV with its suction aligned to the Torus.

CORRECT ANSWER:

D JUSTIFICATION: The specified conditions satisfy the operational requirements for HPCI operation (high DW pressure) only and its suction would automatically align to the Torus once Torus level is greater than +2.

A is incorrect: ADS requires -47" and AC interlock met to initiate B is incorrect: RCIC only auto transfers suction to the Torus at 2'8" in the CST C is incorrect: CS pumps do not inject until ~320 psig

REFERENCE:

B.03.02-02 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

1 CATEGORY:

206000 HPCI K/A:

K4.19 IMPORTANCE:

RO 3.7 COG LEVEL:

2 DR K/A DESCRIPTION:

Knowledge of the HPCI design features and/or interlocks that provide for the following:

Automatic transfer of HPCI pump suction.

DIFFICULTY 2

LESSON PL:

M8107L-002 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

31.

A plant transient has occurred that resulted in an automatic initiation and injection from the Division 1 Core Spray system.

MO-1753 (DIV 1 CS INJECTION INB) is full open with an ECCS initiation signal still present.

Can MO-1753 be used to throttle Core Spray flow? Why or why not?

A. NO, this valve CANNOT be closed with an initiation signal still present.

B. NO, MO-1751 (DIV 1 CS INJECTION OTB) must be used to throttle flow.

C. YES, if HS 14A-S16A (DIV 1 CS INJECTION BYPASS) is first placed in "BYPASS".

D. YES, placing the MO-1753 handswitch in the "CLOSE" position will clear its open signal.

CORRECT ANSWER:

D JUSTIFICATION: After an automatic start of a Core Spray loop, the flow may be throttled with the inboard isolation valve (MO-1753 or MO-1754) using the control switches on Panel C-03. The inboard isolation valve logic is designed such that an automatic open signal is bypassed when its control switch is taken to the close position. The bypassing of the automatic opening signal is then sealed in, regardless of switch position, until the automatic open signal is no longer present.

A is incorrect: The initiation signal doesn't have to be clear and the valve open "seal in" may be bypassed.

B is incorrect: The open signal can only be bypassed by the MO1753 switch, and MO1751 cannot be throttled.

C is incorrect: This switch allows closing of the outboard valve for isolation, not the inboard.

REFERENCE:

B.03.01-01 10 CFR:

55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

1 CATEGORY:

209001 LPCS K/A:

A4.02 IMPORTANCE:

RO 4.2 COG LEVEL:

1I K/A DESCRIPTION: Ability to manually operate and/or monitor the LPCS System in the Control room: Valves.

DIFFICULTY:

3 LESSON PL:

MT-ILT-EOP-002L OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

32.

An ATWS event has occurred with the following resultant conditions:

RPV water level is +20 and slowly lowering

C.5-3101 (ALTERNATE ROD INSERTION) has NOT been performed What will be the status of the following pumps as a result of only placing the SBLC Selector Switch to SYS 1?

Recirc Pumps RWCU Pumps A.

TRIPPED TRIPPED B.

RUNNING TRIPPED C.

TRIPPED RUNNING D.

RUNNING RUNNING CORRECT ANSWER:

B JUSTIFICATION: When a SBLC pump is started by placing the selector switch to SYS 1, the Group 3 isolation logic is initiated causing the running RWCU valves to isolate which causes the RWCU pumps to trip. The Recirc pumps will not automatically trip until RPV water level is < -47 for 7.2 seconds and will not be manually tripped since C.5-3101 has not been performed yet.

A is incorrect: The Recirc pumps will not automatically trip until RPV water level is < -47 for 7.2 seconds and will not be manually tripped since C.5-3101 has not been performed yet.

C is incorrect: The Recirc pumps will not automatically trip until RPV water level is < -47 for 7.2 seconds and will not be manually tripped since C.5-3101 has not been performed yet. When a SBLC pump is started by placing the selector switch to SYS 1, the Group 3 isolation logic is initiated causing the running RWCU valves to isolate which causes the RWCU pumps to trip.

D is incorrect: When a SBLC pump is started by placing the selector switch to SYS 1, the Group 3 isolation logic is initiated causing the running RWCU valves to isolate which causes the RWCU pumps to trip.

REFERENCE:

B.03.05-05.G.1 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

2 GROUP:

1 CATEGORY:

211000 SBLC K/A:

A3.09 IMPORTANCE:

RO 3.8 COG LEVEL:

2 RI K/A DESCRIPTION: Ability to monitor automatic operation of the SBLC system including: Pump trip.

DIFFICULTY 2

LESSON PL:

M8107L-004 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

33.

The reactor was at rated conditions when the following occurred:

A manual scram has been inserted

All scram signals are clear EXCEPT the SDV HI HI level

The SDV HI HI level has been bypassed

The CRS directs the scram to be reset Which position(s) of the Mode Switch will allow the scram to be reset?

A. SHUTDOWN ONLY B. SHUTDOWN OR REFUEL ONLY C. SHUTDOWN OR REFUEL OR START AND HOT STBY ONLY D. SHUTDOWN OR REFUEL OR START AND HOT STBY OR RUN CORRECT ANSWER:

B JUSTIFICATION: The scram signal can only be reset if all scram signals are clear. This can only be accomplished if the mode switch is in either shutdown or refuel and the SDV scram signal is bypassed for this RPS logic arrangement.

A is incorrect: Plausible but not ONLY the shutdown position.

C is incorrect: Low Condenser Vacuum / MSIV Closure scram signal can be bypassed in Shutdown OR Refuel OR Start & Hot Stby. Not the SDV level.

D is incorrect: Plausible but SDV level cannot be bypassed in RUN.

REFERENCE:

B.05.06-02 B.05.06-06 Figure 6 10 CFR 55.41(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2016 NRC Exam #34 TIER:

2 GROUP:

1 CATEGORY:

212000 RPS K/A:

K5.02 IMPORTANCE:

RO 4.1 COG LEVEL:

1 I K/A DESCRIPTION:

Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the RPS System: Logic channel arrangements.

DIFFICULTY 4

LESSON PL:

M8107L-072 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

34.

The Reactor Mode Switch is in STARTUP TO HOT STANDBY. IRM 17 is BYPASSED and ALL other IRM Range Switches are on Range 4. A failure of IRM 11 results in the following IRM readings:

IRM 11 IRM 12 IRM 13 IRM 14 IRM 15 IRM 16 IRM 17 IRM 18 125 70 69 76 72 67 0 68 Based on conditions above; which of the following is the status of RPS and/or RMCS?

A. A Full Scram has occurred.

B. An RMCS rod block ONLY has occurred.

C. An RMCS rod block and a RPS A half scram ONLY have occurred.

D. An RMCS rod block and a RPS B half scram ONLY have occurred.

CORRECT ANSWER:

C JUSTIFICATION: RMCS Rod Block is inserted if one IRM reaches 108. An RPS half scram signal is generated if one RPS A (IRM 11-14) or if one RPS B (IRM 15-18) reaches 119.375. For the conditions above an RMCS Rod Block signal will be generated from IRM 11. An RPS A half scram signal will be generated from IRM 11.

A is incorrect: A full scram will only occur if a half scram signal is generated in both RPS A & B.

B is incorrect: A half scram signal exists in RPS A.

D is incorrect: IRM 17 will not generate a RPS B signal.

REFERENCE:

B.05.06-02 ARP 5-A-13 ARP 5-A-21 10 CFR 55.41b(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

1 CATEGORY:

215003 IRM K/A:

A1.05 IMPORTANCE:

RO 3.9 COG LEVEL:

3 PEO K/A DESCRIPTION:

Ability to predict and/or monitor changes in parameters associated with operation of the IRM System including: Scram and rod block trip setpoints.

DIFFICULTY 2

LESSON PL:

M8107L-072 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

35.

The plant is in a Refuel Outage with the following conditions:

Fuel is being moved into and out of the core

Maintenance is under the vessel staging for CRD Mechanism removal

Shorting links have been REMOVED

IRMs are on Range 1

All SRMs read ~10 cps

Mode Switch is in REFUEL Complete the statement below if the SRM count level was to rise to 5.5 x 105 cps.

Annunciator 5-A-12 (SRM HI/INOP) will...

A. NOT alarm and NO rod block or scram will occur.

B. NOT alarm and ONLY a rod block will occur.

C. alarm and ONLY a rod block will occur.

D. alarm and a FULL scram will occur.

CORRECT ANSWER:

D JUSTIFICATION: Alarm 5-A-12 will alarm at 9.1E4 cps. During fuel movement in and out of the core, the shorting links are removed. With the shorting links removed, any neutron monitoring scram signal (>5.0E5 cps) trips A3 and B3 trip system causing a full scram.

A is incorrect: Plausible for confusion on shorting links. ILT examinees occasionally get mixed up on whether the shorting links are normally removed or installed when shutdown.

B is incorrect: Plausible for not recollecting setpoint and thinking only a half scram will occur.

C is incorrect: Plausible for only thinking a half scram will occur.

REFERENCE:

ARP 5-A-12 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

1 CATEGORY:

215004 SRMs K/A:

K6.01 IMPORTANCE:

RO 3.3 COG LEVEL:

3PEO K/A DESCRIPTION:

Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the SRM System: RPS DIFFICULTY 3

LESSON PL:

M8107L-072 OBJECTIVE:

6

2022 MONTICELLO ILT NRC EXAM - KEY

36.

A plant startup is in progress.

FT-2-110A (REC 11 FLOW XMTR - APRM 1) failed upscale causing the following alarms:

5-A-3 (ROD WITHDRAW BLOCK)

5-A-30 (APRM FLOW BIAS OFF NORMAL)

Complete the statements below?

__(1)__ is preventing control rods from being withdrawn.

Bypassing APRM 1 __(2)__ allow control rods to be withdrawn.

A. (1) RBM Flow Comparison >10%

(2) WILL B. (1) RBM Flow Comparison >10%

(2) WILL NOT C. (1) APRM Flow Upscale >110%

(2) WILL D. (1) APRM Flow Upscale >110%

(2) WILL NOT CORRECT ANSWER:

C JUSTIFICATION: The APRM Flow UPSCALE will cause a rod block for the given conditions. The >10% mismatch is an alarm only. A single FT upscale failure only inputs to APRM 1. If APRM 1 is bypassed the rod block will clear.

A is incorrect: The RBM flow comparison is alarm only.

B is incorrect: Bypassing APRM 1 will allow rods to be withdrawn.

D is incorrect: Plausible dual options based on justification above.

REFERENCE:

ARP 5-A-30 ARP 5-A-3 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2016 NRC Exam #37 TIER:

2 GROUP:

1 CATEGORY:

215005 APRM K/A:

A2.02 IMPORTANCE:

RO 4.1 COG LEVEL:

2 DR K/A DESCRIPTION:

Ability to (a) predict the impacts of the following on the APRM/LPRM and, (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Upscale or downscale trips.

DIFFICULTY 3

LESSON PL:

M8107L-066 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

37.

A plant transient has occurred resulting in an automatic initiation of RCIC.

Shortly after RCIC initiated, flow lowered to 0 gpm and the following indications were observed:

RCIC turbine exhaust pressure is 20 psig

The Mechanical Overspeed light on C-04 is ON

The Group 5 Isolation Reset light on C-04 is ON

MO-2078 (RCIC Turbine Steam Supply Valve) is OPEN

MO-2080 (RCIC Turbine Trip and Throttle Valve) is CLOSED Based on the above indications, what was the cause of the RCIC turbine trip?

A. Mechanical overspeed B. High Reactor water level C. High turbine exhaust pressure D. Group 5 containment isolation CORRECT ANSWER:

A JUSTIFICATION: The mechanical overspeed is designed to trip the RCIC turbine at 5625 rpm in order to prevent turbine damage. This will result in the Mechanical Overspeed light going on C-04 and MO-2080 going closed. All trip signals, except high water level trip the trip throttle valve, MO-2080.

B is incorrect: This would cause MO-2078 to close not MO-2080.

C is incorrect: The high turbine exhaust pressure trip is 50 psig for 5 seconds or 125 psig immediately.

D is incorrect: Group 5 isolation causes a turbine trip and closure of MO-2080 but the Group 5 reset light would be off.

REFERENCE:

B.02.03-01 10 CFR 55.41(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

1 CATEGORY:

217000 RCIC K/A:

A3.07 IMPORTANCE:

RO 4.2 COG LEVEL:

3 SPK K/A DESCRIPTION:

Ability to monitor automatic operation of the RCIC system including: Trips and isolations.

DIFFICULTY 2

LESSON PL:

M8107L-003 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

38.

The plant is at rated conditions when RCIC automatically initiated on a valid initiation signal.

RCIC continues to operate at rated flow.

Which of the following will cause 4-A-14 (RCIC FLOW LOW) to alarm?

A. Torus water level of -5 inches B. Condensate Storage Tank level of 2 feet 5 inches C. CV-2104 (RCIC Pump Minimum Flow Valve) fails open D. RCIC pump suction pressure degrades to 17 inches Hg Vac CORRECT ANSWER:

D JUSTIFICATION: The RCIC Low Flow alarm will annunciate anytime system flow is sensed at = 40 gpm provided MO-2078 (RCIC Turbine Steam Admission Valve) is at least 25% open for 15 seconds or more. A RCIC trip signal will close MO-2080 and leave MO-2078 open which will cause a low flow alarm. RCIC will trip on low suction pressure of 15" Hg Vac and bring in the low flow alarm.

A is incorrect: RCIC does not transfer suction sources on low torus water level. Plausible confusion with HPCI.

B is incorrect: This will automatically transfer RCIC suction source to the torus but the low flow alarm is not expected.

C is incorrect: This will not divert enough flow to cause a low flow alarm.

REFERENCE:

ARP 4-A-14 B.02.03-01 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

2 GROUP:

1 CATEGORY:

217000 RCIC K/A:

K6.13 IMPORTANCE:

RO 3.7 COG LEVEL:

1I K/A DESCRIPTION: Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the RCIC System: Low pump suction pressure DIFFICULTY 2

LESSON PL:

M8107L-003 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

39.

The plant was at rated conditions when a steam line rupture occurred in the drywell and the following conditions are now present:

Drywell Sprays are unavailable

Drywell temperature is 290°F and rising

A complete loss of 125 VDC Bus A has occurred

Emergency Depressurization has been entered

As the BOP, the CRS directs you to open all three ADS valves Given the conditions above, which of the following SRVs must you use to perform the Emergency Depressurization from C-03?

A. SRVs A, C and D.

B. SRV D and two NON-ADS valves.

C. SRV A & C and one NON-ADS valve.

D. ONLY NON-ADS valves will be available.

CORRECT ANSWER:

A JUSTIFICATION: With a loss of 125 VDC Bus A, all power is lost to the A ADS Logic and backup power is lost to the B ADS Logic. 125 VDC Bus B would still be available to actuate ADS with the B ADS Logic. The operator can use all three ADS valves (A, C, & D).

B and C are incorrect: For Emergency depressurization, if an ADS valve doesnt work the operator is directed to use NON-ADS valves. Both of these are plausible for divisional misconceptions and if the examinee doesnt recall that ADS logic power is auctioneered with 125 VDC B.

D is incorrect: Plausible to think none of the valves would work because the use of the LL SET SRVs requires division A power to be opened from C-03.

REFERENCE:

B.03.03-05 B.03.03-06 10 CFR 55.41(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2010 NRC Exam #40 TIER:

2 GROUP:

1 CATEGORY:

218000 ADS K/A:

K2.01 IMPORTANCE:

RO 4.0 COG LEVEL:

2DR K/A DESCRIPTION:

Knowledge of electrical power supplies to the following: ADS logic.

DIFFICULTY 3

LESSON PL:

M8107L-025 OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

40.

The plant is at rated conditions with the HPCI turbine at rated speed for quarterly testing.

HPCI-82 (EXH LINE ISOLATION) experiences a stem failure resulting in a complete blockage of the HPCI exhaust line.

How would this blockage affect the HPCI system if the turbine FAILED TO TRIP on high exhaust pressure, and what action must be taken?

A. The HPCI system is NOT protected in this condition; HPCI must be manually tripped.

B. Rupture discs will burst relieving steam directly to the HPCI room; verify the Group 4 Isolation.

C. Rupture discs will burst relieving steam directly to the Torus catwalk area; verify the Group 4 Isolation.

D. Rupture discs will burst relieving steam directly to the Torus water space; HPCI must be manually tripped.

CORRECT ANSWER:

B JUSTIFICATION: When HPCI turbine exhaust pressure rises to 150 psig the turbine should trip. If a valve stem failure occurs and the turbine fails to trip, two in-line rupture discs will fail at 175 psig relieving the exhaust steam directly to the HPCI Room air space. This will rapidly heat up the HPCI room and cause a Group 4 isolation and automatic HPCI trip when room temperature reaches 187.5°F. The operators would verify the Group 4 isolation.

A is incorrect: The rupture discs will protect the system in this instance.

C is incorrect: The rupture discs do not relieve to the torus water space.

D is incorrect: 8 of the 16 HPCI Group 4 high temperature switches are located in the torus room up near the catwalk area. The rupture discs do not relieve to this room therefore these switches would not activate the Group 4 isolation.

REFERENCE:

B.03.02-01/02 NH-36249 C.4-B.04.01.D 10 CFR 55.41b(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Significantly Modified - 2018 NRC Exam #38 TIER:

2 GROUP:

1 CATEGORY:

223002 PCIS K/A:

A2.07 IMPORTANCE:

RO 3.6 COG LEVEL:

3 PEO K/A DESCRIPTION:

Ability to (a) predict the impacts of the following on the PCIS System and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of the abnormal operations: Various process/instrument failures.

DIFFICULTY 2

LESSON PL:

M8107L-002 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

41.

The plant is at rated conditions when 3-A-17 (AUTO BLOWDOWN VLV BELLOWS LEAKING) alarms and the following indication is observed:

Which of the following describes the ability of SRV A to depressurize the RPV vessel?

SRV A A. will operate automatically in the ADS mode.

B. will operate automatically in the LL-SET mode.

C. will operate for its safety function but at a lower reactor pressure.

D. will NOT operate manually from C-03 to depressurize the RPV vessel.

CORRECT ANSWER:

A JUSTIFICATION: If the bellows is leaking, steam pressure builds up in the pilot stage bonnet. The bellows leaking pressure switches sense the pressure inside the bonnet and are actuated when the pressure builds up to 5 psig. If the sensing bellows should fail, only the safety function (self-actuation) of the valve is disabled (i.e. the self-actuation setpoint of the SRV will drift up by an amount equal to the bellows line pressure.) The valve can still be operated manually or by the ADS system circuitry.

B is incorrect: Plausible if the examinee believes this is a LL-SET SRV.

C is incorrect: It may not operate at all for its safety function and if it does it will be at a higher pressure.

D is incorrect: The manual function is independent of the bellows and will function properly.

REFERENCE:

B.03.03-02 10 CFR 55.41(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2010 NRC Exam #41 TIER:

2 GROUP:

1 CATEGORY:

239002 SRVs K/A:

A4.01 IMPORTANCE:

RO 4.4 COG LEVEL:

3 SPR/SPK K/A DESCRIPTION:

Ability to manually operate and/or monitor the SRVs in the control room: SRVs DIFFICULTY 3

LESSON PL:

M8107L-025 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

42.

The plant was at rated conditions when a 12 Bus Lockout occurred. Immediate Actions have been taken for the trip of 12 Reactor Feed Pump (RFP). Given the following:

CV-3490 (12 RFP FW RECIRC TO CDSR) failed open

11 RFP suction pressure is 92 psig and stable

RPV water level is +27 inches and stable

Reactor power is 62% and stable What action should be taken to mitigate the conditions above?

A. Insert control rods IAW C.4-F (RAPID POWER REDUCTION).

B. Scram the reactor IAW C.4-K (IMMEDIATE REACTOR SHUTDOWN).

C. Raise the setpoint of LC-6-83 (FW MASTER CONTROLLER) back to normal.

D. Raise the setpoint of FC-1095 (11/12 COND PMP RECIRC FLOW) to 4500 gpm.

CORRECT ANSWER:

A JUSTIFICATION: The conditions provided indicate that 11 RFP is operating at or near runout conditions. The procedure directs a power reduction if RFP suction pressure is <100 psig. Reducing power will lower the FW demand and allow RPV level to be restored.

B is incorrect: If reactor power remained above 69% or if reaching +9 were imminent, then the reactor would be scrammed.

C is incorrect: The immediate actions for a RFP trip will lower the setpoint of LC-6-83 to 30. Putting it back to normal

(~38) will not mitigate the situation because the RFP is at max.

D is incorrect: This is an action taken, however the setpoint would be lowered to 4100 gpm.

REFERENCE:

C.4-B.06.05.A 10 CFR 55.41b(10)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2018 NRC Exam #44 - Previous two NRC Exams TIER:

2 GROUP:

1 CATEGORY:

259002 Reactor Water Level Control K/A:

K5.06 IMPORTANCE:

RO 3.2 COG LEVEL:

3 SPK K/A DESCRIPTION:

Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to RWLCS: Pump runout.

DIFFICULTY 4

LESSON PL:

MT-ILT-AOP-015L OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

43.

Given the following conditions:

Drywell pressure is 1.0 psig

Reactor level is 35 inches

"A" SBGT is isolated for maintenance and "B" SBGT is in Standby What is the response of "B" SBGT following a loss of RPS Bus A (Y-50)?

A. Starts immediately.

B. Starts after a 10-second delay.

C. Remains in Standby since no initiation signal is present.

D. Remains in Standby since RPS Bus B (Y-40) is still available.

CORRECT ANSWER:

B JUSTIFICATION: Loss of RPS MG Set "A" will result in a loss of power to the "A" Reactor Building Ventilation Plenum and Fuel Pool Rad Monitors and the "A" trip logic relays. SBGT initiation is provided automatically when:

(A) Following parameters exceed preset limits:

1) Reactor Building ventilation plenum high radiation, one high or two downscale trips of the two Reactor Building exhaust vent plenum monitors (RM-17-452A and RM-17-452B), or

2) Refueling floor radiation, one high or two downscale trips of the two refueling floor process monitors (RM 453A and RM-17-453B)

OR (B) "A" trip logic relays de-energize on loss of RPS "A" If "A" SBGT flow is less than 2800 CFM following a 10 second T.D. from receipt of the initiation signal, the "B" Train will automatically start.

A is incorrect: With SBGT A isolated, a 10 second TD is required for SBGT B to start.

B is incorrect: Plausible as no initiation signal is present, but a loss of power will initiate SBGT.

D is incorrect: Plausible to believe that loss of both RPS busses (RMs) are required initiate SBGT.

REFERENCE:

B.04.02-01 B.05.11-05 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

1 CATEGORY:

261000 SBGT K/A:

K6.01 IMPORTANCE:

RO 3.5 COG LEVEL:

2 RI K/A DESCRIPTION: Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the SBGT System: AC electrical distribution.

DIFFICULTY 3

LESSON PL:

M8107L-008 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

44.

The plant was at rated conditions when a Station Blackout (SBO) occurred.

Which 4160 VAC bus(es) can be procedurally reenergized for the conditions above?

A. Bus 13 ONLY B. Bus 14 ONLY C. Buses 13 and 15 D. Buses 14 and 16 CORRECT ANSWER:

C JUSTIFICATION: Even though 13 DG has restored power to LC-107 and LC-108, this is a plant definition of a Station Blackout (SBO). During a SBO, procedurally, 13 Bus can be reenergized by back feeding from 13 DG. Following that, Bus 15 can be reenergized from Bus 13.

A is incorrect: Bus 15 can also be reenergized.

B is incorrect: Electrically through LC-108, 14 bus could be reenergized and back feeding one bus is plausible but no procedure exists.

D is incorrect: Plausible option as 14 and 16 bus could be reenergized from 13 DG by back feeding via LC-108 but there is no procedure for it and the LC-107/108 crosstie is not rated for both buses.

REFERENCE:

C.4-B.09.02.A E.4-01/03 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

2 GROUP:

1 CATEGORY:

262001 AC Electrical Distribution K/A:

A2.12 IMPORTANCE:

RO 4.6 COG LEVEL:

3 SPR K/A DESCRIPTION: Ability to (a) predict the impacts of the following on the AC Electrical Distribution System; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal operations: Station blackout.

DIFFICULTY 2

LESSON PL:

M8107L-036 OBJECTIVE:

8

2022 MONTICELLO ILT NRC EXAM - KEY

45.

The plant was at rated conditions when SPR-63 (SUDDEN PRESSURE RELAY) actuated in the Main Generator transformer.

Which of the following completes the statements below?

This condition DIRECTLY actuates the __(1)__.

The field breaker trips open __(2)__.

A. (1) turbine lockout relay (286/T)

(2) immediately B. (1) turbine lockout relay (286/T)

(2) once 8N7 and 8N8 are sensed open C. (1) generator lockout relay (286/G)

(2) immediately D. (1) generator lockout relay (286/G)

(2) once 8N7 and 8N8 are sensed open CORRECT ANSWER:

C JUSTIFICATION: (1) Correct, SPR-63; Generator transformer sudden pressure relay will operate for fault in the generator transformer and operation of this relay will cause the generator lockout relay 286/G to trip (2) Correct, generator lockout relay 286/G trips and locks out the field breaker A is incorrect: (1) Incorrect, the turbine lockout relay 286/T is actuated by the generator lockout relay 286/G (2) Correct, generator lockout relay 286/G trips and locks out the field breaker.

B is incorrect: (1) Incorrect, the turbine lockout relay 286/T is actuated by the generator lockout relay 286/G when (2) Incorrect, generator lockout relay 286/G trips and locks out the field breaker. Would be correct for any turbine trip not caused by a generator lockout.

D is incorrect: (1) Correct, SPR-63; Generator transformer sudden pressure will operate for fault in the generator transformer and operation of this relay will cause the generator lockout relay 286/G to trip.

(2) Incorrect, generator lockout relay 286/G trips and locks out the field breaker. Would be correct for any turbine trip not caused by a generator lockout.

REFERENCE:

B.09.02-02 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2016 NRC Exam #47 - Edits to stem TIER:

2 GROUP:

1 CATEGORY:

262001 AC Electrical Distribution K/A:

K4.04 IMPORTANCE:

RO 3.5 COG LEVEL:

2 DR K/A DESCRIPTION: Knowledge of AC Electrical Distribution design features and/or interlocks that provide for the following: Protective relaying.

DIFFICULTY 3

LESSON PL:

M8107L-036 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

46.

Given the following 480 Volt UPS Electrical Distribution Drawing:

An inverter UNDERVOLTAGE fault results in 0 volts sensed at Point X above.

Which of the following describes how Y-94 (480 INST AC PANEL) will receive power?

A. LC-107 will automatically supply power to Y-94.

B. LC-108 will automatically supply power to Y-94.

C. 250 VDC Battery 17 will automatically supply power to Y-94.

D. The Maintenance Bypass Breaker must be manually closed to power Y-94.

CORRECT ANSWER:

A JUSTIFICATION: The examinee must understand that the fault is resulting in a 0 volt output to the static switch.

Therefore, the static switch will automatically transfer to the alternate source (LC-107).

B is incorrect: LC-108 is the normal supply to Y-91 static inverter (via rectifiers) and Battery 17. The inverter is downstream of this and it wouldnt supply power to Y-94.

C is incorrect: The inverter will not transfer to Battery 17 because the battery uses the same input line as LC-108.

D is incorrect: Manual transfer is not required because LC-107 will provide power.

REFERENCE:

B.09.13-01 B.09.13-06 10 CFR 55.41b(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2015 NRC Exam #49 TIER:

2 GROUP:

1 CATEGORY:

262002 UPS K/A:

K4.01 IMPORTANCE:

RO 3.1 COG LEVEL:

3 PEO K/A DESCRIPTION: Knowledge of the UPS System design features and/or interlocks which provide for the following: Transfer of power supplies DIFFICULTY 3

LESSON PL:

M8107L-063 OBJECTIVE:

8

2022 MONTICELLO ILT NRC EXAM - KEY

47.

The plant was at rated conditions when a significant LOCA occurred. Given the following:

ADS has automatically initiated

An RPV depressurization is in progress.

ALL RHR and Core Spray pumps are running Complete the following statement that describes the effect on the ADS System if ALL RHR and Core Spray Pumps TRIP. (Assume RPV water level remains constant at -80 inches and no operator action is taken.)

Automatic Depressurization of the Reactor...

A. continues without Core Spray or RHR pumps running due to the seal-in logic.

B. will cease and re-initiate immediately after a Core Spray or RHR pump is restarted.

C. will cease and re-initiate 107 seconds after a Core Spray or RHR pump is restarted.

D. will cease; both inhibit switches must be taken to INHIBIT and back to AUTO to re-establish depressurization.

CORRECT ANSWER:

C JUSTIFICATION: If the RHR and CS pumps trip, relays K10A and K12A are de-energized, which open up the K10A and K12A contacts to de-energize the 107 second timer, closing the valves. If the pumps are restarted, contacts K10A and K12A close to start the107 second timer again.

A is incorrect: Loss of all ECCS pumps will result in the timer stopping and/or closure of the ADS valves.

B is incorrect: Once the logic resets, the 107 second timer also resets and must time out again.

D is incorrect: Plausible but the inhibit switches need not be operated.

REFERENCE:

B.03.03-01 10 CFR 55.41b(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2015 NRC Exam #43 TIER:

2 GROUP:

1 CATEGORY:

218000 ADS K/A:

K5.01 IMPORTANCE:

RO 4.3 COG LEVEL:

1 I K/A DESCRIPTION:

Knowledge of the operational implications or cause and effect relationships of the following concepts as they apply to the ADS System: ADS logic operation.

DIFFICULTY 3

LESSON PL:

M8107L-025 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

48.

The plant is operating at rated conditions when a DC supply breaker trip results in the receipt of 20-B-09 (DIVISION II 125/250 DC TROUBLE).

Which system must be monitored in the control room to confirm the given power loss?

A. HPCI B. RCIC C. RWCU D. Turbine Lube Oil CORRECT ANSWER:

A JUSTIFICATION: All of the major components for HPCI receive main and/or control power from Division II 250 VDC.

These systems would be monitored in the control room to confirm the power loss.

B is incorrect: RCIC power is receives from D31-10 (Division 1 250 VDC)

C is incorrect: RWCU MO-2398 receives power from D313-09 (Division 1 250 VDC).

D is incorrect: DC powered pumps in this system are powered from the 250 VDC NON-1E 17 Battery.

REFERENCE:

C.4-B.09.09A 10 CFR 55.41(7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

1 CATEGORY:

263000 DC Electrical Distribution K/A:

K2.01 IMPORTANCE:

RO 4.0 COG LEVEL:

1F K/A DESCRIPTION:

Knowledge of electrical power supplies to the following: Major DC loads.

DIFFICULTY 3

LESSON PL:

M8107L-002 OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

49.

The plant is at rated conditions. Given the following:

1AR transformer is tagged out for maintenance.

52-908/CS (109-102 LOAD CENTER TIE ACB 52-908) handswitch in PTL.

If a BUS 13 LOCKOUT were to occur AND 11 EDG failed to start, which Load Centers will be energized ONE MINUTE after the lockout?

LC-101

LC-103

LC-107 A. LC-103 ONLY B. LC-101 AND LC-103 ONLY C. LC-101 AND LC-107 ONLY D. All three Load Centers would be energized CORRECT ANSWER:

C JUSTIFICATION: The 13 bus lockout will trip open the supply breakers to LC-101 and 107. LC-107 would be picked up by #13 Diesel. LC-101 would automatically crosstie to LC-102 since the LC-109 crosstie would not pick it up.

A is incorrect: LC-103 will remain deenergized because 11 EDG failed to start and load onto Bus 15.

B is incorrect: LC-101 will be energized from LC-102 but LC-103 will remain deenergized because 11 EDG failed to start and load onto Bus 15.

D is incorrect: LC-107 will be energized from LC-107 but LC 103 will remain deenergized because 11 EDG failed to start and load onto Bus 15.

REFERENCE:

B.09.07-02 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

1 CATEGORY:

264000 EDGs K/A:

K3.02 IMPORTANCE:

RO 4.4 COG LEVEL:

2 RI K/A DESCRIPTION: Knowledge of the effect that a loss of malfunction of the EDGs will have on the following systems or system parameters: AC Electrical Distribution system.

DIFFICULTY 3

LESSON PL:

M8107L-040 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

50.

A plant startup/heatup is in progress with the following conditions:

CV-2403 (RWCU DUMP FLOW) is 60% open to maintain RPV level

CV-6-13 (LOW FLOW REGULATING VALVE) is closed in Automatic control

A complete loss of Instrument Air occurs While implementing C.4-B.08.04.01.A (LOSS OF INSTUMENT AIR), which of the following valve indications would be expected?

CV-2403 CV-6-13 A.

open closed B.

open open C.

closed open D.

closed closed CORRECT ANSWER:

D JUSTIFICATION: On a loss of IA, CV-2403 will fail closed and CV-6-13 will fail as is.

A is incorrect: CV-2403 will not fail open on a loss of IA.

B is incorrect: CV-2403 will not fail open and CV-6-13 will not fail open.

C is incorrect: CV-6-13 will not fail open.

REFERENCE:

C.4-B.08.04.01.A 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

2 GROUP:

1 CATEGORY:

300000 Instrument Air System K/A:

K1.26 IMPORTANCE:

RO 3.2 COG LEVEL:

2 RI K/A DESCRIPTION: Knowledge of the physical connections and/or cause and effect relationships between the Instrument Air System and the following systems: Reactor Water Cleanup System DIFFICULTY 3

LESSON PL:

M8107L-030 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

51.

A plant startup is in progress with the following conditions:

MO-2399 (RWCU RETURN ISOL) is 25% open

MO-3501 (RBCCW FROM RWCU HX) is 50% open

CV-2403 (RWCU EXCESS FLOW CTRL VLV) is controlling dump flow at 60 gpm With the conditions above which of the following actions must be taken if annunciator 4-B-26 (CLEANUP DEMIN TEMP HI) is received?

A. Throttle OPEN CV-2403.

B. Throttle OPEN MO-3501.

C. Throttle CLOSED MO-3501.

D. Throttle CLOSED MO-2399.

CORRECT ANSWER:

B JUSTIFICATION: The examinee must determine that this is an alarm only at 125°F. Opening MO-3501 will send more RBCCW to the RWCU HX thus lowering temperature. If this action isnt taken IAW the ARP, temperature will continue to rise and a RWCU Hi temp isolation will occur at 140°F causing MO-2399 to close.

A is incorrect: Throttling this valve open will cause temperature to rise more.

C is incorrect: The ARP procedure gives the option to manipulate this valve but it must be opened to lower temperature.

D is incorrect: Closing this valve will further limit the return flow through the NRHX causing the filter demin temp to rise further.

REFERENCE:

ARP 4-B-26 10 CFR 55.41b(10)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2015 NRC Exam #53 TIER:

2 GROUP:

1 CATEGORY:

400000 Component Cooling Water K/A:

2.4.12 IMPORTANCE:

RO 4.0 COG LEVEL:

3 SPK K/A DESCRIPTION: Knowledge of operating crew responsibilities during emergency and/or abnormal operations.

DIFFICULTY 3

LESSON PL:

M8107L-030 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

52.

The normal Service Water supply has been lost to the following coolers:

V-AC-6 (RCIC ROOM COOLER)

V-AC-7A/B (CRD ROOM COOLERS)

V-AC-8A/B (HPCI ROOM COOLERS)

If EFT-ESW Pumps were started, which cooler(s), if any, will still be monitored for outlet radiation?

[Service Water System P&IDs NH-36041 (M-110) and NH-36664 (M-112) are provided]

A. V-AC-8A/B ONLY B. V-AC-8A/B and V-AC-6 ONLY C. V-AC-7A/B and V-AC-6 and ONLY D. NO coolers will be monitored for outlet radiation.

CORRECT ANSWER:

A JUSTIFICATION: HPCI Room Coolers V-AC-8A/B is the only room coolers listed above that would be cooled by ESW in the event of a loss of normal service water. See P&ID M-112 (C,3). With this cooler still having flow through it, the discharge would still be monitored by the Service water radiation monitor. See P&ID M-112 (A, 6).

B is incorrect: V-AC-6 is not supplied by ESW however it is unique as its the only cooler directly monitored by the Discharge Canal Radiation monitoring system.

C is incorrect: V-AC-7A/B is not supplied by ESW however, its plausible as CRD and HPCI are both located on the NW side of the RB.

D is incorrect: V-AC-8A/B is the only cooler supplied by ESW.

REFERENCE:

B.08.01.01-01 10 CFR 55.41b(12)

REFERENCE PROVIDED DURING EXAM:

P&IDs NH-36041 and NH-36664 QUESTION SOURCE:

New TIER:

2 GROUP:

1 CATEGORY:

510000 Service Water System K/A:

K3.10 IMPORTANCE:

RO 2.5 COG LEVEL:

3 SPR K/A DESCRIPTION:

Knowledge of the effect that a loss or malfunction of the Service Water System will have on the following systems or system parameters: Radiation Monitoring System.

DIFFICULTY 3

LESSON PL:

M8107L-077 OBJECTIVE:

2.j

2022 MONTICELLO ILT NRC EXAM - KEY

53.

The plant was at rated conditions when an event occurred resulting in the following readings:

1. Control Room Air Intake Monitors - 2.5 mrem/hr

2. Refueling Floor Radiation Monitors - 60 mrem/hr

3. RB Exhaust Plenum Radiation Monitors - 21 mrem/hr Which of the given instrument radiation readings will cause CRV/EFT to initiate in the High Radiation Mode?

A. 1 ONLY B. 2 ONLY C. 1 & 2 ONLY D. 1, 2 AND 3 CORRECT ANSWER:

C JUSTIFICATION: Control Room Air Intake monitors will initiate the High Radiation Mode at 1 mrem/hr and the Refueling Floor Radiation monitors will initiate High Radiation Mode at 50 mrem/hr.

A is incorrect: Prior to the modification during the 2007 RFO, this would have been the correct answer. This modification added 2 and 3 to the initiation logic.

B is incorrect: Plausible if the examinee thinks the Control room intake radiation monitors initiates at a radiation higher than 2.5 mrem/hr or if examinee confuses High Rad Mode with standard secondary containment isolation.

D is incorrect: The RB Exhaust Plenum Radiation monitors are reading above the alarm setpoint (20 mrem/hr) but they are not above the initiation setpoint of 26 mrem/hr.

REFERENCE:

B.08.13-01 10 CFR 55.41(11)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2010 NRC Exam #18 - Edits to stem TIER:

2 GROUP:

2 CATEGORY:

290003 Control Room Ventilation K/A:

K1.01 IMPORTANCE:

RO 3.5 COG LEVEL:

1I K/A DESCRIPTION: Knowledge of the physical connections and/or cause and effect relationships between the CRV System and the following systems: Radiation Monitoring System.

DIFFICULTY 2

LESSON PL:

M8107L-049 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

54.

Given the following:

Control rod 26-27 is at position 48

Control rod 30-19 is at position 48 and is presently being inserted to position 42 using normal rod insertion Complete the statement below?

Annunciator 5-A-27, ROD DRIFT, will alarm _______________.

A. if rod 26-27 moves off position 48 B. when rod 30-19 moves past position 47 C. if the TIMER TEST switch is taken to the RESET position D. when the ROD MOVEMENT CONTROL switch is released to the OFF position CORRECT ANSWER:

A JUSTIFICATION: Setpoint for Rod Drift: Lack of CLOSED even numbered reed switch or occurrence of CLOSED odd number reed switch in the Position Indicator Probe of any Control Rod which is not selected and driving.

B is incorrect: The drift alarm will not come in if that rod is selected and being moved.

C is incorrect: This will cause a drift alarm if the switch is taken to TEST.

D is incorrect: This will not cause a drift alarm unless another malfunction prohibited the proper settle function.

REFERENCE:

ARP 5-A-27 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

2 CATEGORY:

214000 RPIS K/A:

K4.04 IMPORTANCE:

RO 3.9 COG LEVEL:

1I K/A DESCRIPTION: Knowledge of RPIS design features and/or interlocks that provide for the following: Detection of a drifting control rod.

DIFFICULTY 2

LESSON PL:

M8107L-032 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

55.

Which of the following will lose power on a loss of D11 (DIVISION 1 125 VDC)?

A. TIP Shear Valve B. SRMs 21 and 22 C. IRMs 15, 16, 17 and 18 D. HPCI System Control Power CORRECT ANSWER:

A JUSTIFICATION: D11, Ckt 15 supplies power to TIP Control Cabinet C-13 Shear Valve control.

B is incorrect: Plausible VDC supply but this is D15 (DIVISION 1 24 VDC)

C is incorrect: Plausible VDC supply but this is D25 (DIVISION 2 24 VDC)

D is incorrect: Plausible VDC supply but this is D100 (DIVISION 2 125 VDC). 250 VDC center tapped.

REFERENCE:

B.05.03-05 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

2 GROUP:

2 CATEGORY:

215001 TIP K/A:

K2.01 IMPORTANCE:

RO 3.1 COG LEVEL:

1F K/A DESCRIPTION: Knowledge of electrical power supplies to the following: Shear Valves.

DIFFICULTY 2

LESSON PL:

M8107L-059 OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

56.

The plant was at rated conditions when a Torus break and LOCA occurred. The following conditions are observed:

RPV pressure is 25 psig

Torus water level is -6 feet

RHR pumps are unavailable

DW/Torus pressure is 8 psig and slowly rising

RHRSW-14 (EMERG INJ VIA A RHRSW) will NOT open

P&IDs M-112 and M-121 for RHRSW and ESW Systems are provided Which of the following methods can be procedurally used to spray the Torus?

A. RHRSW B. Fire Water C. Domestic Well Water D. Condensate Service Water CORRECT ANSWER:

D JUSTIFICATION: With the low torus water level, RHR pumps would be unavailable to run. This would require alternate methods to be used for Torus Spray. With the conditions and options listed above, Condensate Service water would be the only option to spray the Torus.

A is incorrect: This would be an option if RHRSW-14 could open.

B is incorrect: This would be an option if RHRSW-14 could open.

C is incorrect: There is currently no procedure available to spray the Torus with domestic well water.

REFERENCE:

C.5-3502 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

P&IDs NH-36664(M-112) NH-36247 (M-121)

QUESTION SOURCE:

Bank - Edits to stem TIER:

2 GROUP:

2 CATEGORY:

230000 RHR Torus Spray Mode K/A:

K6.12 IMPORTANCE:

RO 3.7 COG LEVEL:

2 RI K/A DESCRIPTION: Knowledge of the effect of the following plant conditions, system malfunctions, or component malfunctions on the RHR Torus Spray Mode: Low suppression pool level.

DIFFICULTY 2

LESSON PL:

MT-ILT-EOP-003L OBJECTIVE:

3

2022 MONTICELLO ILT NRC EXAM - KEY

57.

The plant is in a refuel outage with the following conditions established:

Reactor Cavity Flooded

Fuel Pool to Cavity Gates Removed If the reactor cavity bellows were to develop an unisolable leak, which of the following represents how low water level would get in the fuel pool with no operator action?

(Assume emergency makeup is NOT available.)

A. ONLY the upper tie plates of the fuel would become UNCOVERED.

B. ONLY 1 foot of the fuel would become UNCOVERED.

C. ONLY 1/3 of the fuel would become UNCOVERED.

D. ONLY 2/3 of the fuel would become UNCOVERED.

CORRECT ANSWER:

A JUSTIFICATION: The FPC&C response procedure Rapid Loss of Spent Fuel Pool Water Level Through the Reactor Cavity Bellows Seals states the following precaution: Loss of fuel pool water level through the reactor cavity bellows seals is possible only when the reactor cavity is flooded and the fuel pool to cavity gates are open. Fuel handling activities may or may not be in progress. In such a scenario, it is possible for water level to drop nearly as low as the top of the spent fuel storage racks. This leaves the upper tie plates of the stored fuel uncovered, but the top of the active fuel is approximately one foot under water. All fuel should be adequately cooled, but abnormal radiation levels may exist.

B is incorrect: Plausible misconception of one foot above vs. one foot below.

C is incorrect: Plausible level to ensure melting of the fuel doesnt occur.

D is incorrect: Plausible confusion with 2/3rds core height.

REFERENCE:

B.02.01-05.H.7 10 CFR 55.41(13)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2010 NRC Exam #10 TIER:

2 GROUP:

2 CATEGORY:

233000 FPC&C K/A:

A1.02 IMPORTANCE:

RO 3.6 COG LEVEL:

1P K/A DESCRIPTION: Ability to predict and/or monitor changes in parameters associated with operation of FPC&C including: Fuel pool level.

DIFFICULTY 3

LESSON PL:

M8107L-022 OBJECTIVE:

9

2022 MONTICELLO ILT NRC EXAM - KEY

58.

Which plant conditions meet the definition of REFUELING MODE 5?

A. Reactor water temperature is 100°F Mode switch is in STARTUP/HOT STANDBY RPV head is removed B. Reactor water temperature is 180°F Mode switch is in SHUTDOWN RPV head is removed C. Reactor water temperature is 200°F Mode switch is in REFUEL ALL RPV closure head bolts fully tensioned D. Reactor water temperature is 221°F Mode switch is in SHUTDOWN ALL RPV closure head bolts fully tensioned CORRECT ANSWER:

B JUSTIFICATION: It is the only answer that meets the definition of Mode 5.

A is incorrect: This would be Mode 2.

C is incorrect: This would be Mode 2.

D is incorrect: This would be Mode 3.

REFERENCE:

TS Table 1.1-1 Above 10 CFR 55.41b(10)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Significantly Modified - 2018 NRC Exam #70 TIER:

2 GROUP:

2 CATEGORY:

234000 Fuel Handling K/A:

A4.03 IMPORTANCE:

RO 3.7 COG LEVEL:

1 D K/A DESCRIPTION: Ability to manually operate and/or monitor the Fuel Handling in the control room: Mode switch.

DIFFICULTY 2

LESSON PL:

MT-OPS-ITS-002L OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

59.

With the plant operating at full power, the indicated flow on the "D" Main Steam Line went fully upscale. Actual steam flow rose to 123% flow for two seconds. The flow indications on the "A", "B" and "C" Main Steam Lines lowered to less than their normal 100% flow during the two second transient.

Which statement describes expected plant protection response to this condition?

A. Group I isolation resulting in automatic reactor scram.

B. Trip of Main Steam Line High Flow Channel B ONLY.

C. No response until the high flow is sensed for > 5 seconds.

D. No response until high flow is sensed on a 2nd main steam line.

CORRECT ANSWER:

A JUSTIFICATION: The logic is arranged so any high flow (>116.9%) in a single steam line will cause isolation. A high flow condition on any one main steam line will result in a Group I isolation.

B is incorrect: Both channels will receive a trip.

C is incorrect: There is no time delay for this trip.

D is incorrect: Only one steam line required.

REFERENCE:

B.05.06-02, C.6-5-A-52 10 CFR 55.41 (7)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

2 CATEGORY:

239001 Main and Reheat Steam K/A:

A1.09 IMPORTANCE:

RO 3.7 COG LEVEL:

3 PEO K/A DESCRIPTION: Ability to predict and/or monitor changes in parameters associated with operation of the Main and Reheat Steam system including: Main steam flow.

DIFFICULTY 2

LESSON PL:

M8107L-070 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

60.

The plant was operating at rated conditions in a normal electric plant lineup. An event occurs that results in the following:

Annunciator 20-A-08 (FIRE) is alarming

Annunciator 8-A-01 (NO. 1 GENERATOR LOCKOUT) is alarming

The Main Transformer Fire System status is shown below The BOP operator rotates the arming collar and depresses the pushbutton shown above IAW ARP 20-A-08.

Which of the following, if any, is a valid reason for performing these actions?

A. These actions MANUALLY INITIATE deluge spray on the Main Generator Transformer.

B. These actions MANUALLY TRIP open the 345 KV Main Generator output breakers 8N7 and 8N8.

C. These actions MANUALLY TRIP open the Main Generator Field breaker which results in a Generator LOCKOUT.

D. These actions are NOT required; the Main Transformer Deluge IS currently spraying water on the transformer.

CORRECT ANSWER:

A JUSTIFICATION: The deluge for the Main XFMR will not automatically spray when the LHD detects a fire. The Detector Operated light indicates that a fire has been detected which auto opens the deluge inlet valve which fills the piping with water. In order to actually spray water, the Main XFMR must be verified dead (Main Generator LOCKOUT) then the arming collar must be turned and the pushbutton pushed. This action is called for in the Abnormal procedure for FIRE and ARP 20-A-08.

B and C are incorrect: The Main Generator Lockout will automatically trip these breakers. Plausible that this action must be manually performed for a fire.

D is incorrect: Plausible to think this action isnt required because the turbine building siding deluge automatically sprays when the detector is operated. The Main XFMR deluge spray doesnt auto initiate as indicated by the System Operated light being off.

REFERENCE:

ARP 20-A-08 C.4-B.08.05.A 10 CFR 55.41b(10)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2015 NRC Exam #75, 2016 NRC Exam #19 TIER:

2 GROUP:

2 CATEGORY:

245000 Main Turbine Generator & Aux K/A:

2.4.25 IMPORTANCE:

RO 3.3 COG LEVEL:

2 DR K/A DESCRIPTION: Knowledge of fire protection procedures.

DIFFICULTY 3

LESSON PL:

MT-ILT-AOP-016L OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

61.

The plant is in MODE 1 with a shutdown in progress. Given the following:

Main Turbine Bypass Valves (BV) are closed

LP Turbine exhaust hood temperature is 120°F and stable

CV-1269 (COND HDR TO TURB EXH HOOD SPRAY) is in AUTO

Main Turbine Control Valves (CVs) are open approximately 12% of rated position If condensate flow is lost through CV-1269, which temperature, if any, will FIRST initiate annunciator 7-B-06/07 (LP TURB G-1A/B EXH HOOD HIGH TEMP)?

A. 175°F B. 200°F C. 225°F D. None, this alarm is inactive with CVs <15% open.

CORRECT ANSWER:

A JUSTIFICATION: TS-1258 will initiate alarm 7-B-6 when exhaust hood temperatures reach 175°F.

B is incorrect: This is the temperature that a power reduction may be required IAW 7-B-06.

C is incorrect: This is the temperature that a scram may be required IAW 7-B-06.

D is incorrect: Plausible for confusion with the auto open signal to CV-1269 once CVs are <15% open.

REFERENCE:

ARP 7-B-6 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

2 GROUP:

2 CATEGORY:

256000 Condensate System K/A:

K3.14 IMPORTANCE:

RO 2.5 COG LEVEL:

1 I K/A DESCRIPTION: Knowledge of the effect that a loss or malfunction of the Condensate System will have on the following systems or system parameters: Exhaust hood spray system DIFFICULTY 3

LESSON PL:

M8107L-013 OBJECTIVE:

9

2022 MONTICELLO ILT NRC EXAM - KEY

62.

Which of the following is correct concerning the Cable Spreading Room (CSR) fire dampers?

CSR fire dampers A. must be closed manually from the ASDS Panel.

B. must be closed manually from PAB 2 outside the CSR.

C. automatically close from Halon pressure in the supply lines.

D. automatically close when control is transferred to the ASDS Panel.

CORRECT ANSWER:

C JUSTIFICATION: Automatic closure of Cable Spreading Room fire dampers is accomplished by Halon pressure in the supply lines.

A is incorrect: Plausible location as plant control would like be transferred to ASDS.

B is incorrect: Plausible location to take manual action.

D is incorrect: Plausible automatic trigger for the damper closure.

REFERENCE:

B.08.05-01 B.08.05-05 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

2 GROUP:

2 CATEGORY:

286000 Fire Protection System K/A:

A3.06 IMPORTANCE:

RO 2.8 COG LEVEL:

1I K/A DESCRIPTION: Ability to monitor automatic operation of the Fire Protection System including: Fire Dampers.

DIFFICULTY 3

LESSON PL:

M8107L-010 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

63.

The plant is at 50% power with the Circulating Water System in PARTIAL RECIRC.

Both CWPs are running

11 CTP is running

The CWP/CT Pump Auto Trip Interlock Bypass Switch is in AUTO.

A Circ Water rupture results in the receipt of 101-A-11 (HOTWELL AREA FLOODING) and water level on the Pit Floor is currently at 3 feet.

What will be the status of the CWPs and 11 CTP following this event?

11 CWP 12 CWP 11 CTP A.

RUNNING RUNNING RUNNING B.

TRIPPED RUNNING TRIPPED C.

TRIPPED TRIPPED RUNNING D.

TRIPPED TRIPPED TRIPPED CORRECT ANSWER:

D JUSTIFICATION: In the Condenser Pit area if the water level continues to rise to Elev 909'-6" (1 Foot) and two of the three level switches are tripped...... the Circulating Water pumps will trip.

With both CWPs and the 11 CTP running and the Second CWP trips, The 11 CTP will trip when in partial recirc.

A is incorrect: Plausible for not knowing the correct setpoint for hotwell area flooding.

B is incorrect: Plausible for not knowing the auto CWP actions for hotwell area flooding.

C is incorrect: Plausible for not knowing the auto CTP actions for hotwell area flooding.

REFERENCE:

B.06.04-02 10 CFR 55.41b(2)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

2 GROUP:

2 CATEGORY:

510001 Circ Water System K/A:

K5.03 IMPORTANCE:

RO 3.3 COG LEVEL:

1 I K/A DESCRIPTION: Knowledge of the operational implications and or cause and effect relationships of the following concepts as they apply to the Circ Water System: Pipe rupture.

DIFFICULTY 3

LESSON PL:

M8107L-017 OBJECTIVE:

7

2022 MONTICELLO ILT NRC EXAM - KEY

64.

Which of the following requires the assigned individual(s) to be respirator qualified?

1. Being a Fire Brigade member

2. Being a Control Room Operator during a Toxic Gas event

3. Being an ASDS Panel Operator for Shutdown Outside the Control Room

4. Being a Control Room Operator during ALL Emergency Plan implementations A. 1 and 2 ONLY B. 1, 2 and 3 ONLY C. 1, 2 and 4 ONLY D. 1, 2, 3, and 4 CORRECT ANSWER:

A JUSTIFICATION: Fire Brigade and Toxic gas events are on-shift collateral duties that require respirator qualifications.

B is incorrect: Plausible for fire in the control room, but not required.

C is incorrect: Plausible for high radiation events, but not required.

D is incorrect: Plausible as all situations listed above could require a respirator but the duties of 1 and 2 are the only ones that require it for qualifications.

REFERENCE:

OWI-01-06 10 CFR 55.41b(10)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2016 NRC Exam #74 TIER:

3 GROUP:

1 CATEGORY:

Conduct of Operations K/A:

2.1.2 IMPORTANCE

RO 4.1 COG LEVEL:

1 P K/A DESCRIPTION: Knowledge of operator responsibilities during any mode of plant operation.

DIFFICULTY 3

LESSON PL:

M8108L-038 OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

65.

Complete the statement below that indicates a loss of the 125 Vdc Bus A has occurred.

The white power lights below the annunciator panels will turn OFF on Control Room Panels A. C-03, C-04, and C-05 ONLY B. C-06, C-07, and C-08 ONLY C. C-03, C-04, C-05, C20, and C-259 D. C-06, C-07, C-08, C20, and C-259 CORRECT ANSWER:

B JUSTIFICATION: These annunciator panels are powered from 125 Vdc bus A and the white lights are power available light which turn off when there is a loss of this power.

A is incorrect: C-03/04/05 are powered from 125 VDC Bus B.

C is incorrect: C-03/04/05/20 and C-259 are powered from 125 VDC Bus B D is incorrect: Plausible divisional confusion with the extra panels but C-20 and C-259 are powered from 125 VDC Bus B

REFERENCE:

C.4-B.05.13A Tables 1 and 2 10 CFR 55.41b(7)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

3 GROUP:

1 CATEGORY:

Conduct of Operations K/A:

2.1.19 IMPORTANCE:

RO 3.9 COG LEVEL:

1 F K/A DESCRIPTION: Ability to use available indications to evaluate system or component status.

DIFFICULTY 3

LESSON PL:

M8107L-025 OBJECTIVE:

6

2022 MONTICELLO ILT NRC EXAM - KEY

66.

A plant startup is in progress with RPV pressure at 800 psig and both Recirc loops in service.

Which of the following by itself would be a Safety Limit violation with the above conditions?

A. MCPR changed to 0.99 B. MCPR changed to 1.12 C. Reactor power rose to 27%

D. Rated core flow lowered to 8%

CORRECT ANSWER:

A JUSTIFICATION: With reactor steam dome pressure > 586 psig and core flow >10% rated core flow MCPR shall be

>1.05.

B is incorrect: Plausible as the TS MCPR safety limit recently changed from must be > 1.15. Also plausible for confusion between MFLCPR which is not good above 1.

C is incorrect: With reactor steam dome pressure < 586 psig OR <10% rated core flow then THERMAL POWER shall be

< 25% RTP. With both Recirc loops in service, Recirc flow will be at least 30%.

D is incorrect: Being < 10% rated core flow by itself is not a safety limit violation.

REFERENCE:

TS 2.1.1.3 10 CFR 55.41b(10)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

3 GROUP:

2 CATEGORY:

Equipment Control K/A:

2.2.22 IMPORTANCE:

RO 3.1 COG LEVEL:

1 B K/A DESCRIPTION: Knowledge of limiting conditions for operation and safety limits DIFFICULTY 3

LESSON PL:

MT-ILT-OPS-003L OBJECTIVE:

1

2022 MONTICELLO ILT NRC EXAM - KEY

67.

The plant is at rated conditions with all electrical buses on their normal supply and the Subyard aligned as follows for maintenance:

10 Bank Transformer is ISOLATED

Breaker 1N2 (13.8 KV OCB) is OPEN

Breaker 1N6 (13.8 KV OCB) is OPEN A 1R Transformer LOCKOUT now occurs with the conditions above.

Are Tech Spec LCO requirements met for electrical power requirements? Why or Why not?

A. YES, the plant currently has sufficient Qualified Offsite Circuits.

B. YES, the plant currently has sufficient NSP Transmission Lines.

C. NO, the plant currently does NOT have sufficient Qualified Offsite Circuits.

D. NO, the plant currently does NOT have sufficient NSP Transmission Lines.

CORRECT ANSWER:

C JUSTIFICATION: 2R Transformer is the normal supply to plant busses.

Qualified Offsite Circuits required by TS 3.8.1 are as follows (two of the four are required):

1) 2R Transformer

2) 1R Transformer

3) 1AR Transformer fed from 10 Bank via 1N2(normally) or

4) 1AR Transformer fed from 1ARS via 1N6 2R Transformer is only available; therefore, TS 3.8.1 Condition A must be entered.

Qualified NSP Transmission Lines (two of the six are required) are as follows:

345 KV - Elm Creek Substation, Sherburne County Substation and Quarry Substation.

115 KV - Hassan, Dickenson/Lake Pulaski, and Liberty Subyard.

All of the 345 KV transmission lines are available, 3 of 6 are available so TLCO 3.8.1 is met.

A is incorrect: Plausible if examinee believes 1AR is still available. Tech Spec entry would be required for lack of offsite circuits.

B is incorrect: The required transmission lines are met but a Tech Spec entry is required for lack of offsite circuits.

D is incorrect: The required transmission lines are met. If they werent this would be a TRM entry, not Tech Specs.

REFERENCE:

TS 3.8.1, TRM 3.8.1 and B.09.03-05 10 CFR 55.41(7, 10)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

3 GROUP:

2 CATEGORY:

Equipment Control K/A:

2.2.42 IMPORTANCE:

RO 3.9 COG LEVEL:

2RI K/A DESCRIPTION: Ability to recognize system parameters that are entry-level conditions for technical specifications.

DIFFICULTY 3

LESSON PL:

M8114L-003 OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

68.

Which of the following is correct concerning the plant Area Radiation Monitors (ARMs)?

A. ARM channels must be observed twice per shift.

B. Plant page announcements should be made when trip testing ARMs.

C. ARM trip and indicator units are powered from Division I and II 24 VDC.

D. If an ARM is taken out-of-service for >1 hour, RP must periodically check radiation levels in that area.

CORRECT ANSWER:

B JUSTIFICATION: When trip testing ARMs, it should be announced over the Plant Public Address System. This should be done to prevent unwarranted local and site evacuations.

A is incorrect: ARMs are required to be observed daily.

C is incorrect: All ARMs are powered from a single power supply. At no time shall power be turned off. The power is from Instrument AC Panel Y-20. Plausible as Process Radiation Monitors are powered from 24 VDC.

D is incorrect: RP must periodically check radiation level in the area if an ARM is out-of-service for more than 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

REFERENCE:

B.5.12-05 Plant Operating Requirements and General Precautions 10 CFR 55.41b(11)

REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

3 GROUP:

3 CATEGORY:

Radiation Control K/A:

2.3.5 IMPORTANCE

RO 2.9 COG LEVEL:

1 P K/A DESCRIPTION:

Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

DIFFICULTY 3

LESSON PL:

M8107L-077 OBJECTIVE:

9

2022 MONTICELLO ILT NRC EXAM - KEY

69.

A Site Area Emergency has been declared at the plant. You are currently offsite and have been called in to report to the Operations Support Center (OSC).

Which building is the OSC located within?

A. Monticello Training Center B. Plant Engineering Building C. Plant Admin Building D. Warehouse 1 CORRECT ANSWER:

C JUSTIFICATION: The OSC is located in various areas of the first and second floors of the PAB.

A is incorrect: Plausible location as this is where the EOF is located.

B is incorrect: A back-up (alternate) OSC (BOSC) is located in the TSC on the first level of the Plant Engineering Building (PEB).

D is incorrect: Plausible location as this in located in between the PAB and the PEB.

REFERENCE:

A.2-107 10 CFR 55.41b(10)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

3 GROUP:

4 CATEGORY:

Emergency Procedures/Plan K/A:

2.4.42 IMPORTANCE:

RO 2.6 COG LEVEL:

1 F K/A DESCRIPTION: Knowledge of emergency response facilities.

DIFFICULTY 2

LESSON PL:

MT-BEP-OPS-001L OBJECTIVE:

5

2022 MONTICELLO ILT NRC EXAM - KEY

70.

The plant is at steady-state rated conditions.

Which of the following indicate that thermal hydraulic oscillations are occurring? (Evaluate each condition separately)

A. APRM recorders indicating a peak-to peak oscillation of 1.5%.

B. Core thermal power at 30% with Total Core flow at 30 Mlbm/hr.

C. A single LPRM UPSCALE alarm in the center region of the core.

D. RBM ODAs indicating a peak-to-peak LPRM oscillation of 12.0%.

CORRECT ANSWER:

D JUSTIFICATION: LPRM bargraphs can be displayed on the RBM or APRM ODAs and will show overall core flux distribution. LPRMs typically indicate some random oscillation. The steady state LPRM noise level should be within + 5%

of scale. LPRM thermal flux oscillations of 12% would meet the entry indications for Abnormal Operating Procedure for Control of Neutron Flux Oscillations.

A is incorrect: 10% peak-to-peak would meet the entry indications.

B is incorrect: At this power and flow the OPRM region of the P-F Map would be entered but isnt an indication of flux oscillations.

C is incorrect: Multiple LPRM upscale alarms in a wide area of the core that repeatedly alarm and then clear would be an indication.

REFERENCE:

C.4-B.05.01.02.A 10 CFR:

55.41b(5)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

4 GROUP:

292001 CATEGORY:

Reactor Theory K/A:

K1.08 IMPORTANCE:

RO 2.4 COG LEVEL:

1P K/A DESCRIPTION: Describe fast flux, thermal flux and flux distribution.

IR<2.5 has been noted and approved for use by the Chief Examiner pending the ability to develop an operationally valid question associated with core flux.

DIFFICULTY:

2 LESSON PL:

MT-ILT-AOP-002L OBJECTIVE:

6

2022 MONTICELLO ILT NRC EXAM - KEY

71.

Prior to a refueling outage, all or nearly all control rods are fully withdrawn at 100% power.

After a refueling outage, a significant amount of control rods are inserted much farther into the core at 100% power.

What is the reason for the difference in full power control rod position?

(EOC = End of Cycle, BOC = Beginning of Cycle)

A. Reactivity from power defect at BOC is much greater than at EOC.

B. The integral control rod worth at EOC is much greater than at BOC.

C. Reactivity from void coefficient at EOC is much greater than at BOC.

D. The excess reactivity in the core at BOC is much greater than at EOC.

CORRECT ANSWER:

D JUSTIFICATION: K-excess or excess reactivity is excess fuel that is added to core beyond minimum amount necessary to achieve criticality at BOL. Therefore, less control rods need to be fully withdrawn at BOC. At the EOC excess reactivity is gone an all control rods will eventually be withdrawn during coast down.

A is incorrect: Power defect reactivity is lower at BOC.

B is incorrect: Integral rod worth at EOC is lower.

C is incorrect: Void coefficient reactivity is lower at EOC.

REFERENCE:

Generic Fundamentals 10 CFR 55.41b(1)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

4 GROUP:

292002 CATEGORY:

Reactor Theory K/A:

K1.09 IMPORTANCE:

RO 2.6 COG LEVEL:

1 F K/A DESCRIPTION: Define K-excess (excess reactivity)

DIFFICULTY 2

LESSON PL:

M8120L-120 OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

72.

The plant was at 70% power for 11 days when power was raised to 100% in a 2-hour period.

Once at 100% power, what minor Recirc flow adjustments will be made to offset the change in Xenon-135 concentration over the next 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months ?

A. Lower Recirc flow due to Xe-135 burnout.

B. Raise Recirc flow due to Xe-135 burnout.

C. Lower Recirc flow due to I-135 decay.

D. Raise Recirc flow due to I-135 decay.

CORRECT ANSWER:

A JUSTIFICATION: When power is raised, Xenon concentration will initially lower due to removal by burnout. This will add positive reactivity requiring recirc flow to be lowered. This will last for ~ 5 hours5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months then it will rise from the decay of Iodine requiring recirc flow to be raised.

B is incorrect: Initially recirc flow will have to be lowered to compensate from xenon burnout.

C is incorrect: Plausible but xenon concentration will lower due to burnout.

D is incorrect: Plausible but recirc flow will need to be lowered due to xenon burnout.

REFERENCE:

Generic Fundamentals 10 CFR 55.41b(1)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

New TIER:

4 GROUP:

292006 CATEGORY:

Reactor Theory K/A:

K1.04 IMPORTANCE:

RO 2.9 COG LEVEL:

2 DR K/A DESCRIPTION: Describe the removal of xenon-135.

DIFFICULTY 3

LESSON PL:

M8120L-124 OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

73.

The plant was at rated conditions when a scram occurred. RPV pressure has been stable at 925 psig for one hour.

Which of the following is the MINIMUM pressure that the reactor may be reduced to at the end of a one-hour cooldown without violating Technical Specifications?

A. 405 psig B. 385 psig C. 365 psig D. 345 psig CORRECT ANSWER:

C JUSTIFICATION: The maximum allowed cooldown is 100°F/hr when averaged over a one hour period. Exceeding this limit will likely risk damage to the RPV and internal components. See data below:

Pressure Pressure Saturation Temp Delta Temp (PSIG) (PSIA) (°F) (°F) 925 940 537.197 N/A 405 420 449.431 87.77 385 400 444.533 92.66 365 380 439.634 97.56 345 360 434.321 102.88 A is incorrect: This would be allowed but not the minimum.

B is incorrect: Plausible for steam table misapplication of psig vs. psia.

D is incorrect: Plausible options for incorrect application of steam table pressure and temperature tables.

REFERENCE:

TS 3.4.9 PTLR 10 CFR 55.41b(3,

14)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

Steam Tables and Mollier Diagram QUESTION SOURCE:

New TIER:

4 GROUP:

293003 CATEGORY:

Thermodynamics K/A:

K1.23 IMPORTANCE:

RO 3.1 COG LEVEL:

3 SPR K/A DESCRIPTION:

Use saturated and superheated steam tables DIFFICULTY 3

LESSON PL:

M8107L-028 OBJECTIVE:

8

2022 MONTICELLO ILT NRC EXAM - KEY

74.

The plant has been at rated power for several weeks. It has been determined that air binding is occurring in the main condenser waterboxes.

Which of the following would result from the air binding?

A. Offgas system flow will rise.

B. Main Generator output will rise.

C. Condenser vacuum will degrade.

D. Circ Water condenser outlet temperature will lower.

CORRECT ANSWER:

C JUSTIFICATION: Air binding in the condenser water boxes will cause lower circ water flow and this will cause main condenser vacuum to degrade. Understanding the condensing process is required to know the effects of air binding.

A is incorrect: With degraded condenser vacuum, there is less vacuum to draw in non-condensables and this will cause Offgas flow to lower.

B is incorrect: With degraded condenser vacuum, main generator output will lower as the generator is operating less efficiently.

D is incorrect: Circ water condenser outlet temperature will rise if one or more waterboxes are getting air bound.

REFERENCE:

C.4-B.06.04.A 10 CFR 55.41b(5, 14)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank - 2009 NRC Exam #21 TIER:

4 GROUP:

293004 CATEGORY:

Thermodynamics K/A:

K1.14 IMPORTANCE:

RO 2.7 COG LEVEL:

2 RI K/A DESCRIPTION: Explain the condensing process.

DIFFICULTY 2

LESSON PL:

MT-ILT-AOP-014L OBJECTIVE:

12

2022 MONTICELLO ILT NRC EXAM - KEY

75.

A rapid rise in the differential temperature (DT) between the fuel clad and the coolant with a lowering in heat flux indicates _________.

A. bulk boiling is occurring B. nucleate boiling is occurring C. critical heat flux (CHF) is rising D. departure from nucleate boiling (DNB)

CORRECT ANSWER:

D JUSTIFICATION: DNB occurs when steam bubbles coalesce and form a vapor film along the heat transfer surface.

This occurs at the point of maximum heat flux. This results in a rapid rise in differential temperature and a lowering of heat transfer.

A is incorrect: Bulk boiling occurs when the bulk of the flowing fluid is at saturation temperature for system pressure.

B is incorrect: Nucleate boiling occurs when steam bubbles form in a liquid along a heat transfer surface.

C is incorrect: CHF may have been reached resulting in DNB, but this value is constant at this point.

REFERENCE:

Generic Fundamentals 10 CFR 55.41b(14)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None QUESTION SOURCE:

Bank TIER:

4 GROUP:

293008 CATEGORY:

Thermodynamics K/A:

K1.08 IMPORTANCE:

RO 3.1 COG LEVEL:

1 F K/A DESCRIPTION: Describe departure from nucleate boiling.

DIFFICULTY 3

LESSON PL:

M8120L-116 OBJECTIVE:

10

2022 MONTICELLO ILT NRC EXAM - KEY

76.

The plant is in MODE 5 during a refueling outage with the following conditions:

All SRVs are gagged CLOSED

The Reactor is fueled with the Cavity flooded

Fuel Pool Gates are IN with a drain time of > 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months

Decay Heat load is within the RWCU capacity.

The Condensate & Feedwater System is unavailable

Division 2 AC and DC electrical buses are de-energized for emergent maintenance

11 RHR pump and 13 RHRSW pump are in Shutdown Cooling An electrical fault results in an 11 RHR Pump LOCKOUT.

What will be the resultant Decay Heat Removal risk color and what action should be taken regarding the Division 2 electrical bus maintenance? (Form 2270 pg. 2 is provided on the following page).

A. YELLOW; stop the maintenance until 11 RHR pump is restored.

B. YELLOW; continue the maintenance around the clock until completed.

C. ORANGE; stop the maintenance until 11 RHR pump is restored.

D. ORANGE; continue the maintenance around the clock until completed.

CORRECT ANSWER:

D JUSTIFICATION: The loss of 11 RHR pump will result in a loss of forced core flow circulation. DHR will be orange based on only two subsystems available (13 RHR pump and 11 RWCU pump). Equipment maintenance, either voluntary or emergent, resulting in an ORANGE condition should be worked to completion around the clock to minimize time in this condition.

A is incorrect: Plausible if examinee doesnt realize CS cant be counted for DHR without SRVs and thinks maintenance should be stopped.

B is incorrect: Plausible if examinee doesnt realize CS can be counted for IC without SRVs.

C is incorrect: Plausible if examinee thinks maintenance should be stopped.

REFERENCE:

2270 FP-OP-ROM-02 pg. 41 10 CFR:

55.43b(1)

EXTERNAL REFERENCE PROVIDED DURING EXAM:

Embedded Form 2270 pg. 2 SRO ONLY JUSTIFICATION:

SRO Only task to perform Shutdown Risk Assessments.

QUESTION SOURCE:

New TIER:

1 GROUP:

1 CATEGORY:

295001 Partial or Complete loss of forced core flow circulation.

K/A:

2.2.18 IMPORTANCE:

SRO 3.9 COG LEVEL:

3 SPR K/A DESCRIPTION: Knowledge of the process for managing maintenance activities during shutdown operations such as risk assessments and work prioritization.

DIFFICULTY:

3 LESSON PL:

MT-ILT-AOP-007L OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

77.

The plant was at rated conditions when a severe fire started in the Cable Spreading Room.

Given the following:

Immediate Control Room evacuation was successfully completed

Main Turbine Bypass Valves have failed OPEN

MSIVs failed to automatically close

RPV pressure is lowering What should be directed to the ASDS Panel operator and performed by the CRS?

A. Close the INBOARD MSIVs and declare an NUE.

B. Close the OUTBOARD MSIVs and declare an NUE.

C. Close the INBOARD MSIVs and declare an ALERT.

D. Close the OUTBOARD MSIVs and declare an ALERT.

CORRECT ANSWER:

D JUSTIFICATION: IF the RPV rapidly depressurizes OR RPV water level cannot be maintained, THEN place the MSIV closure switch to CLOSE. ONLY the outboard MSIVs can be closed from the ASDS Panel. The higher ALERT classification would be declared due to transfer of plant control to the ASDS Panel (HA6.1)

A is incorrect: Plausible if examinee believes the Inboard MSIVs can be closed from the ASDS Panel. Additionally, plausible as an NUE could be declared for the fire but the Alert would be the higher classification.

B is incorrect: Plausible as an NUE could be declared for the fire but the Alert would be the higher classification.

C is incorrect: Plausible if examinee believes the Inboard MSIVs can be closed from the ASDS Panel.

REFERENCE:

C.4-C B.05.17-02 EAL Matrix 10 CFR:

55.43b(1, 5)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

EAL Board (5790-101-02)

SRO ONLY JUSTIFICATION:

Abnormal action determination and EAL declaration QUESTION SOURCE:

ILT Bank - 2018 Audit Exam TIER:

1 GROUP:

1 CATEGORY:

295016 Control Room Abandonment K/A:

AA2.03 IMPORTANCE:

SRO 4.2 COG LEVEL:

3 SPK/SPR K/A DESCRIPTION: Ability to determine or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Reactor Pressure DIFFICULTY:

3 LESSON PL:

MT-ILT-AOP-025L OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

78.

The plant is at rated conditions when a Train A N2 leak results in the following:

PI-4895 (ALT N2 TRAIN A HEADER PRESSURE) reads 85 psig

AI-719 (ALT N2 TRAIN A PCV-4904 ISOL) is OPEN Which of the following is correct?

A. ONLY Declare one LLS valve inoperable Restore ALT N2 Train A pressure within 14 days.

B. ONLY Declare one ADS valve inoperable Restore ALT N2 Train A pressure within 14 days.

C. Declare one ADS valve AND one LLS valve inoperable Restore ALT N2 Train A pressure within 14 days.

D. Declare two ADS valves inoperable Place the plant in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

CORRECT ANSWER:

C JUSTIFICATION: Operability Requirements: Train A supplies one ADS SRV (A), one Lo-Lo Set SRV (E), and B SRV.

Train B supplies one ADS SRV (C), one Lo-Lo Set SRV (H), and F SRV. Trains A and B must be operable when ADS and Lo-Lo Set functions of the SRVs are required to be operable. If one Train of the Alternate Nitrogen System is inoperable, the ADS capability of one SRV becomes inoperable, and the Lo-Lo Set capability of one SRV becomes inoperable. The appropriate ADS and Lo-Lo Set Technical Specification LCOs would be entered at that point. IAW SR 3.5.1.3, N2 bank header pressure must be > 88.3 psig for ADS actuation operability. One ADS valve would need to be restored within 14 days.

A is incorrect: One ADS valve would also be inoperable.

B is incorrect: One LLS valve would also be inoperable.

D is incorrect: Only one ADS valve would be affected by N2. Plausible required action for lack of system knowledge.

REFERENCE:

B.08.04.03-05 TS 3.5.1 TS 3.6.1.5 10 CFR:

55.43b(2)

EXTERNAL REFERENCE PROVIDED DURING EXAM:

TS 3.5.1, 3.6.1.5 SRO ONLY JUSTIFICATION:

Tech Spec Action determination.

QUESTION SOURCE:

Bank - Edits to stem and choices.

TIER:

1 GROUP:

1 CATEGORY:

295019 Partial or Complete loss of Instrument Air K/A:

AA2.01 IMPORTANCE:

SRO 3.9 COG LEVEL:

3 SPR K/A DESCRIPTION: Ability to determine or interpret the following as they apply to Partial or Complete loss of Instrument Air: Instrument Air Pressure.

Note: Instrument Air and Instrument N2 systems are physically connected and separate from Alt N2. Additionally, this is recent OE at MNGP with high operational validity.

DIFFICULTY:

3 LESSON PL:

M8107L-024 OBJECTIVE:

9

2022 MONTICELLO ILT NRC EXAM - KEY

79.

The plant is in MODE 5 when a Refueling Accident occurred at 1230. Given the following:

At 0100 the Stack WRGMs read as follows:

Stack Effluent Monitor Channels A & B read 5.0E+5 µCi/Sec.

At 0105 the Stack WRGMs read as follows:

Stack Effluent Monitor Channel A & B read 6.8E+5 µCi/Sec.

Of the times below, when is the EARLIEST that conditions will be met to declare an ALERT?

(Assume radiation levels continue to rise at the same rate)

A. 0100 B. 0105 C. 0110 D. 0115 CORRECT ANSWER:

C JUSTIFICATION: STACK ALERT LEVEL: 8 E+5. The CRS must interpret the rate of change and determine EAL RA1.1 for WRGM values for declaring an ALERT will be met 0110 (8.6 E+5).

A is incorrect: Plausible as the NUE level has been reached by 0100.

B is incorrect: Plausible as readings have increased but are still not at Alert level.

D is incorrect: Alert conditions have been met but this is not the earliest time.

REFERENCE:

EAL Matrix 10 CFR:

55.43b(1, 5)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

EAL Board (5790-101-02)

SRO ONLY JUSTIFICATION:

EAL Matrix Interpretation QUESTION SOURCE:

Significantly Modified - 2015 NRC Exam #93 TIER:

1 GROUP:

1 CATEGORY:

295023 Refueling Accidents K/A:

AA2.05 IMPORTANCE:

SRO 4.4 COG LEVEL:

3 SPR K/A DESCRIPTION: Ability to determine or interpret the following as they apply to REFUELING ACCIDENTS:

Emergency Plan Implementation.

DIFFICULTY:

3 LESSON PL:

MT-OPS-BEP-01 OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

80.

The plant is at rated conditions when alarm 5-B-28 (DRYWELL HI PRESS SCRAM TRIP) is received. AOP C.4-B.04.01.F (LEAK INSIDE PRIMARY CONTAINMENT) was entered but an investigation determined the alarm was due to a faulty pressure switch.

Troubleshooting is performed by Maintenance and provides you with the following data:

RPS Drywell Pressure - High Inst. Number As-Found Setting (PSIG)

PS-5-12A 0.52 PS-5-12B 1.95 PS-5-12C 2.05 PS-5-12D 2.09 Based on this information, which of the following, if any, is the most limiting Technical Specification REQUIRED ACTION and COMPLETION TIME?

A. Restore RPS trip capability within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

B. Place one trip system in trip within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

C. Place the channel(s) in trip within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

D. None, all pressures are within the ALLOWABLE VALUE.

CORRECT ANSWER:

B JUSTIFICATION: The low as-found setting for 12A caused the alarm but the allowable value is less than or equal to 2 psig. Based on the given information the C (A2) and the D (B2) instruments are outside this value. This meets Condition B which states One or more Functions with one or more required channels inoperable in both trip systems.

A is incorrect: This would only be correct if the A&C or the B&D instruments were out of tolerance.

C is incorrect: This would only be correct if only one instrument was out of tolerance.

D is incorrect: All as found values would have to be less than or equal 2 psig.

REFERENCE:

TS 3.3.1.1 10 CFR 55.43b(2)

REFERENCE PROVIDED DURING EXAM:

TS 3.3.1.1 SRO ONLY JUSTIFICATION Technical Specification action determination QUESTION SOURCE:

Bank TIER:

1 GROUP:

1 CATEGORY:

295024 High Drywell Pressure K/A:

2.2.36 IMPORTANCE:

SRO 4.2 COG LEVEL:

3 SPR K/A DESCRIPTION: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

DIFFICULTY 3

LESSON PL:

M8108L-039 OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

81.

The plant is at rated condition in a normal electrical lineup with the following conditions:

Reactor power indicates 2003 MWth

The EPR Control Position is set at 904 PSI

The MPR Handwheel Position indicates 914 PSI An equipment malfunction causes the MCC-142A supply breaker to TRIP OPEN.

For the above conditions:

1) Which of the following procedures must be directed?

2) What is a correct reason for entering this procedure?

A. 1) C.4-K (IMMEDIATE REACTOR SHUTDOWN)

2) To lower reactor power below the license limit.

B. 1) C.4-K (IMMEDIATE REACTOR SHUTDOWN)

2) For a Pressure Regulator setpoint failing downscale.

C. 1) C.4-B.05.09.B (PRESSURE REGULATOR FAILURE CAUSING INCREASED PRESSURE)

2) To lower reactor pressure in order to lower reactor power below the license limit.

D. 1) C.4-B.05.09.B (PRESSURE REGULATOR FAILURE CAUSING INCREASED PRESSURE)

2) For a Pressure Regulator setpoint failing downscale.

CORRECT ANSWER:

C JUSTIFICATION: Both EPR oil pumps A & B are powered by MCC-142A, so a loss will cause the MPR to take control at a reactor pressure approximately 10 psig higher. General precautions; During power operation, changing the controlling pressure regulator setpoint will have a small but noticeable effect on core thermal power (approximately 3MWt per 10 psig change). The examinee must apply this thumb rule and know the higher reactor pressure will cause power to rise to 2006 MWth which would require a power reduction to remain within the license limit (2004 MWth).

A is incorrect: This procedure isnt used to perform power reductions.

B is incorrect: This procedure isnt used to perform power reductions; however, this option is plausible for misunderstanding how the EPR will fail. The EPR pressure regulator setpoint will fail upscale with the loss of MCC-142. If a regulator failed downscale this would be a correct procedure entry.

D is incorrect: The EPR pressure regulator setpoint will fail upscale with the loss of MCC-142. Plausible for confusing required actions for regulator failure upscale or downscale.

REFERENCE:

C.4-B.09.07.D C.4-B.05.09.B B.05.09-05 10 CFR:

55.43b(5)

REFERENCE PROVIDED DURING EXAM:

None SRO ONLY JUSTIFICATION:

Assessment of conditions and selection of appropriate procedure.

QUESTION SOURCE:

Bank - 2015 NRC Exam #85 TIER:

1 GROUP:

1 CATEGORY:

295025 High Reactor Pressure K/A:

EA2.02 IMPORTANCE:

SRO 4.2 COG LEVEL:

2 DR K/A DESCRIPTION: Ability to determine or interpret the following as they apply to HIGH REACTOR PRESSURE:

Reactor Power DIFFICULTY:

3 LESSON PL:

MT-ILT-AOP-021 OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

82.

The plant is at rated conditions:

4-B-4 (SUPPRESSION WATER LEVEL HI/LOW) is received

Torus water level indicates -0.4 feet on SPDS Complete the statements below:

(1) IAW the Technical Specification Bases, are the DBA LOCA assumed initial conditions satisfied for indicated Torus water level?

(2) If the Torus Downcomer lines become uncovered, which capability will be lost?

A. (1) satisfied (2) pressure suppression function B. (1)

NOT satisfied (2) pressure suppression function C. (1) satisfied (2)

Torus cooling capability D. (1)

NOT satisfied (2)

Torus cooling capability CORRECT ANSWER:

B JUSTIFICATION: DBA LOCA conditions assume that Torus water level is above -4.0. The analysis in C.5-1-1001 states that when the downcomers are uncovered then pressure suppression capability is lost and primary containment pressure could exceed structural limits. 4-B-4 is received at -2 however the examinee must interpret the level indication of - 0.4 feet on SPDS actually correlates to -4.8, therefore the DBA initial conditions are NOT satisfied.

A is incorrect: Plausible if the trainee does not understand the basis for staying above -4.0 in the Torus.

C is incorrect: Plausible if the trainee believes that Torus cooling capability is the limiting factor for uncovering the downcomers.

D is incorrect: Plausible if the trainee both doesnt understand the basis for -4.0 and believes that loss of Torus cooling capability is the limiting factor for uncovering the downcomers.

REFERENCE:

TS 3.6.2.2 Bases 10 CFR:

55.43b(2)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None SRO ONLY JUSTIFICATION:

Knowledge of Tech Spec Bases QUESTION SOURCE:

Bank - 2016 NRC Exam #81 - Edits to stem TIER:

1 GROUP:

1 CATEGORY:

295030 Low Supp. Pool Water Level K/A:

2.1.45 IMPORTANCE:

SRO 43 COG LEVEL:

2 DR K/A DESCRIPTION: Ability to identify and interpret diverse indication to validate the response of another indication.

DIFFICULTY:

2 LESSON PL:

MT-ILT-EOP-003L OBJECTIVE:

5

2022 MONTICELLO ILT NRC EXAM - KEY

83.

A transient occurred resulting in the following conditions:

Drywell pressure is 58 psig and slowly rising

Torus water level is 13 ft and slowly rising

Drywell radiation is 10 R/hr and slowly rising

Attempts to spray the drywell have been unsuccessful

Significant fuel damage is NOT anticipated You have decided to perform C.5-3505 (VENTING PRIMARY CONTAINMENT).

Which one of the following choices identifies:

(1) The recommended vent path?

(2) The desired strategy for venting Primary Containment?

A. (1) SBGT through the 18 inch torus vent (C.5-3505 PART C)

(2) Venting MUST be limited to ONLY the volume required to maintain pressure below the DW pressure limit B. (1) SBGT through the 18 inch torus vent (C.5-3505 PART C)

(2) Venting MAY be extended for a period of time to reduce the amount of radioactivity that may have to be released once fuel damage occurs C. (1) Hard Pipe Vent (C.5-3505 PART A)

(2) Venting MUST be limited to ONLY the volume required to maintain pressure below the DW pressure limit D. (1) Hard Pipe Vent (C.5-3505 PART A)

(2) Venting MAY be extended for a period of time to reduce the amount of radioactivity that may have to be released if fuel damage occurs CORRECT ANSWER:

C JUSTIFICATION: The preferred method for venting PC is through the Torus using SBGT However, if SBGT ductwork is in jeopardy of rupturing due to high DW pressure (>2.9 psig) and Torus water level >11.3, then the HPV should be used. With significant fuel damage NOT expected (DW Rads @ 10R), venting MUST be limited to ONLY the volume required to maintain pressure below the DW pressure limit.

A is incorrect: Dont vent through SBGT due to potential impacts on RB ductwork and Torus water level being > 11.3.

B is incorrect: Extended venting would only be performed if significant fuel damage was expected. Venting would not be performed through SBGT due to potential impacts on RB ductwork and Torus water level being > 11.3.

D is incorrect: Extended venting would only be performed if significant fuel damage was expected.

REFERENCE:

C.5-3505 C.5.1-1001 10 CFR:

55.43b(5)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None SRO ONLY JUSTIFICATION:

Procedure selection during Emergency.

QUESTION SOURCE:

Significantly Modified - 2016 NRC Exam #83 TIER:

1 GROUP:

2 CATEGORY:

295010 High DW Pressure K/A:

AA2.03 IMPORTANCE:

SRO 3.6 COG LEVEL:

3 SPR K/A DESCRIPTION:

Ability to determine/interpret the following as it applies to High DW Pressure: Drywell Rad Levels DIFFICULTY:

3 LESSON PL:

MT-ILT-EOP-003L OBJECTIVE:

3

2022 MONTICELLO ILT NRC EXAM - KEY

84.

The plant was at rated conditions when a seismic event occurred.

The following conditions exist:

RPV water level is -60 and slowly rising

RPV pressure is 600 psig and slowly lowering

HPCI is injecting at rated flow

Both trains of Drywell Sprays are in operation.

Torus level is +2.5 feet and stable

CST level is +3 feet and stable Given the above conditions, which of the following Torus Level mitigating actions should be prioritized by the CRS?

A. Drain Torus Water to Radwaste.

B. Stop RPV injection from the HPCI system.

C. Perform Emergency RPV Depressurization.

D. Stop Drywell Sprays and anticipate RPV Depressurization.

CORRECT ANSWER:

A JUSTIFICATION: Restore and maintain torus level below +3.0 inches Efforts to restore torus water level to the normal band should be continued while the need for further action is evaluated. The CRS should prioritize supplemental procedure C.5-3402 (DRAINING TORUS WATER TO RADWASTE). This procedure provides instructions for reducing torus water level by transferring water to Radwaste through RHR. If torus water level cannot be restored and maintained below 3.7 feet, a blowdown will ultimately be required. Torus water level is above 3 but is relatively stable at 3 and being maintained below 3.7.

B is incorrect: Torus level has exceeded the swap over level to align the HPCI suction to the torus, therefore HPCI is no longer taking suction from a source outside the Containment.

C is incorrect: Torus water level is currently being maintained below 3.7, therefore Blowdown in not required.

D is incorrect: Torus water level currently being maintained at approximately 3, therefore level is well below the threshold (4.2) for the specified action.

REFERENCE:

C.5-1200 C.5.1-1001 C.5-3402 10 CFR:

55.43b(5)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

C.5-1200 SRO ONLY JUSTIFICATION:

EOP Action Determination QUESTION SOURCE:

Bank - 2016 NRC Exam #85 TIER:

1 GROUP:

2 CATEGORY:

295029 High Torus Water Level K/A:

2.4.16 IMPORTANCE:

SRO 4.4 COG LEVEL:

3 SPR K/A DESCRIPTION: Knowledge of emergency and abnormal operating procedures implementation hierarchy and coordination with other support procedures or guidelines such as, operation procedures, abnormal operating procedures or sever accident management guidelines.

DIFFICULTY:

2 LESSON PL:

MT-ILT-EOP-003L OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

85.

The plant was at rated conditions when a steam leak occurred in the RWCU room from RC-3-1 (11 RWCU PUMP SUCT) requiring a reactor scram. Given the following:

C.4-B.02.04.A (STEAM LEAKS OUTSIDE PRIMARY CONT) has been entered

The RWCU system automatically isolated from a Group 3 Isolation

RWCU Room entry must be made to close and DANGER tag RC-3-1

RWCU Room radiation levels and temperatures remain elevated Can the CRS waive Independent Verification (IV) to close and tag RC-3-1, why or why not?

A. NO, DANGER tag IV cannot be waived on safety related valve RC-3-1.

B. NO, the position of RC-3-1 can be IVed on Control Room Panel C-04.

C. YES, DANGER tag IV can be waived for the room conditions above.

D. YES, DANGER tag IV is only required in MODE 1 operation.

CORRECT ANSWER:

C JUSTIFICATION: Independent verification for Danger tags may be waived by the CRS when significant radiation, safety or other hazards exist. This valve is not safety related but is located in the RWCU Pump Room. This room is a locked high radiation area within secondary containment.

A is incorrect: It can be waived and RC-3-1 is not safety related.

B is incorrect: In some cases, IV can be performed with remote indications; however, RC-3-1 does not have a remote indication on control room panel C-04.

D is incorrect: Danger tag IVs apply to all modes of operation.

REFERENCE:

4 AWI-04.04.02 Section 4.2.9 AOP C.4-B.02.04.A 10 CFR:

55.43b(1, 4)

EXTERNAL REFERENCE PROVIDED DURING EXAM:

None SRO ONLY JUSTIFICATION:

SRO Only task to approve waiving of independent verification.

QUESTION SOURCE:

New TIER:

1 GROUP:

2 CATEGORY:

295032 High Secondary Containment Area Temperature K/A:

2.2.13 IMPORTANCE:

SRO 4.3 COG LEVEL:

1 P K/A DESCRIPTION: Knowledge of tagging and clearance procedures DIFFICULTY:

2 LESSON PL:

M8108L-039 OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

86.

The plant was at normal rated conditions with the #12 RHR pump operating in Torus Cooling Mode when a LOCA occurred resulting in the following indications:

Time RPV Level RPV Pressure Drywell pressure Torus Temperature 1004

-50 inches 1020 psig 1.0 psig 75°F 1021

-55 inches 800 psig 1.8 psig 78°F 1023

-60 inches 600 psig 1.9 psig 83°F 1025

-70 inches 700 psig 2.0 psig 90°F

Emergency Depressurization will NOT be required and ADS actuation was inhibited.

Based on the above conditions, complete the statements below:

At 1021 the #12 RHR pump __(1)__ be operating in the Torus Cooling Mode.

At 1025 you should direct Control Room personnel to __(2)__.

A. (1) will (2) Prevent ALL injection from CS and LPCI B. (1) will (2) Transfer A and B RHR from LPCI to Torus Cooling C. (1) will NOT (2) Prevent ALL injection from CS and LPCI D. (1) will NOT (2) Transfer A and B RHR from LPCI to Torus Cooling CORRECT ANSWER:

D JUSTIFICATION: A LPCI Initiation Signal on Low-Low reactor water level for 15 minutes becomes active at 1019, this sends a signal to close MO-2007 and MO-2009 [torus cooling valves] so a path for torus cooling is no longer aligned. After the ECCS actuation, Torus temperature begins to rise at an increasing rate and at 1025 Torus Temperature has reached a value requiring use of all available Torus Cooling. EOP C.5-1100 also directs that CS and LPCI injection not needed for core cooling be prevented. The examinee must balance the need for increased injection and the need for Torus Cooling.

A is incorrect: (1) MO-2007 and 2009 received a closed signal at 1019, the flow path for torus cooling is no longer aligned. (2) With RPV level continuing to decrease additional injection is needed; preventing injection from all low pressure ECCS sources would not be appropriate.

B is incorrect: (1) MO-2007 and 2009 received a closed signal at 1019; the flow path for torus cooling is no longer aligned.

C is incorrect: (2) With RPV level continuing to decrease additional injection is needed; preventing injection from all low pressure ECCS sources would not be appropriate.

REFERENCE:

C.5-1100/1200 B.03.04-02/05 10 CFR:

55.43b(5)

REFERENCE PROVIDED DURING EXAM:

EOP-1100/1200 with entry conditions blanked out SRO ONLY JUSTIFICATION:

EOP action determination.

QUESTION SOURCE:

Bank - 2016 NRC Exam #91 - Edits to stem TIER:

2 GROUP:

1 CATEGORY:

203000 RHR/LPCI Injection mode K/A:

A2.16 IMPORTANCE:

SRO 4.4 COG LEVEL:

3 SPR/SPK K/A DESCRIPTION:

Ability to (a) predict the impacts of the following on the RHR/LPCI Injection Mode and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: LOCA DIFFICULTY:

2 LESSON PL:

M-BEP-OPS-001L OBJECTIVE:

1

2022 MONTICELLO ILT NRC EXAM - KEY

87.

The reactor was operating at 30% power. I&C technicians were performing calibration testing on the Group 1 isolation logic and inadvertently caused a Group 1 isolation. Reactor pressure peaked at 1052 psig.

Given FP-OP-REP-01 Attachment 2 (10CFR 50.72 REPORTABLE PLANT EVENTS); which is the most restrictive 10CFR 50.72 notification required, if any?

A. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B. 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C. 8 hour9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months D. None required CORRECT ANSWER:

B JUSTIFICATION: With the Group 1 isolation a reactor scram (RPS Actuation) will have occurred. This a 4-hour notification IAW 50.72(b)(2)(iv)(B).

A is incorrect: Entry into E plan is not required and a deviation of plant Technical Specifications has not occurred C is incorrect: These conditions fall under the 8-hour notification, however the RPS actuation is more restrictive.

D is incorrect: Plausible if candidate doesnt realize a RPS trip still occurs even though reactor pressure only reached 1052 psig (scram set point is 1056 psig).

REFERENCE:

FP-OP-REP-01 Attachment 2 NUREG 1022 10CFR50.72 10 CFR:

55.43(1)

REFERENCE PROVIDED DURING EXAM:

FP-OP-REP-01 Attachment 2 SRO ONLY JUSTIFICATION:

Immediate notification requirements.

QUESTION SOURCE:

Bank - 2009 NRC Exam #81 TIER:

2 GROUP:

1 CATEGORY:

212000 RPS K/A:

2.4.30 IMPORTANCE:

SRO 4.1 COG LEVEL:

3SPR K/A DESCRIPTION:

Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the state, the NRC, or transmission system operator.

DIFFICULTY:

2 LESSON PL:

M8108L-039 OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

88.

The plant was at rated conditions when the following SRV A indications were observed:

C.4-B.03.03.A (STUCK OPEN RELIEF VALVE) is in progress

SRV A handswitch was placed in OPEN and RETURNED to the normal position

Reactor power was reduced and is currently stable at 80%

SRV A indication remains as indicated above Which C.4-B.03.03.A action must be directed next by the CRS?

A. Bypass the Div II Lo-Lo Set Logic on CR panel C-253D.

B. Remove the four fuses associated with SRV A on CSR panel C-32.

C. Place the SRV A handswitch in the CLOSE position on CR panel C-03.

D. Place the SRV A handswitch in the CLOSE position on ASDS panel C-292.

CORRECT ANSWER:

B JUSTIFICATION: In the picture SRV A (ADS) is shown to be opened from an inadvertent ADS signal. ADS valves are only on C-03 and have two switch positions (AUTO-OPEN). The stuck open SRV procedure states to place the SRV in OPEN then back to AUTO. If that doesnt close the valve then reactor power is reduced. If the SRV remains open then the CRS must direct an operator to remove fuses on CSR panel C-32.

A is incorrect: This action would be taken for Lo-Lo Set SRVs (E, F, G, H).

C is incorrect: ADS valves dont have a CLOSE position, only Lo-Lo Set valves do.

D is incorrect: Only Lo-Lo SET valves can be operated from ASDS.

REFERENCE:

C.4-B.03.03.A 10 CFR 55.43b(5)

REFERENCE PROVIDED DURING EXAM:

None SRO ONLY JUSTIFICATION:

Abnormal procedural action requires CRS direction to remove fuses to disable an ADS valve. Question selected as discussed with Chief Examiner during outline development to match K/A.

QUESTION SOURCE:

New TIER:

2 GROUP:

1 CATEGORY:

218000 ADS K/A:

2.1.9 IMPORTANCE

SRO 4.5 COG LEVEL:

3 SPR/SPK K/A DESCRIPTION:

Ability to direct licensed personnel activities inside the control room.

DIFFICULTY 3

LESSON PL:

MT-ILT-AOP-006L OBJECTIVE:

4

2022 MONTICELLO ILT NRC EXAM - KEY

89.

The plant is in MODE 1 when annunciator 5-A-46 (SRV OPEN) is received along with the following indications:

Which of the following is a required TRM/Tech Spec REQUIRED ACTION and COMPLETION TIME for the above conditions?

A. None; TRM/Tech Spec Actions are NOT required.

B. Restore SRV E to OPERABLE status within 14 days.

C. Restore the instrument channel to OPERABLE status within 30 days D. Restore the instrument channel to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months .

CORRECT ANSWER:

C JUSTIFICATION: The SRO examinee must determine if the SRV OPEN alarm is consistent with conditions shown in the pictures. The amber light lit means pressure switch has activated (30#), which would indicate that an S/RV is open; however, the associated MSL flow does not reflect that the S/RV is open and tailpipe temperature is consistent with the SRV being closed. This is indicative of a pressure switch instrument failure which requires entry into condition A.

A is incorrect: Plausible if the examinee does not think a TS or TRM is applicable to this pressure switch.

B is incorrect: Plausible if the examinee believes that the associated LLSET valve in INOPERABLE, however, since no indication exists that anything is wrong with the valve itself, it should not be assumed INOPERABLE.

D is incorrect: Plausible if the examinee believes that the associated S/RV is a Low-Low Set valve.

REFERENCE:

TRM TLCO 3.3.3.1 10 CFR:

55.43b(2)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

TLCO 3.3.3.1, & TS 3.3.6.3, 3.4.3, 3.5.1, 3.6.1.5 SRO ONLY JUSTIFICATION:

TS Action determination.

QUESTION SOURCE:

Bank - 2013 NRC Exam #100 - Edits to stem TIER:

2 GROUP:

1 CATEGORY:

239002 SRVs K/A:

2.2.45 IMPORTANCE:

SRO 4.7 COG LEVEL:

3 SPR/SPK K/A DESCRIPTION: Ability to determine or interpret technical specifications with action statements > 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

DIFFICULTY:

3 LESSON PL:

M8107L-025 OBJECTIVE:

10

2022 MONTICELLO ILT NRC EXAM - KEY

90.

The plant was at rated conditions with 11 EDG operating in parallel with Off-Site power IAW 0187-01 (11 EDG/ESW QUARTERLY PUMP AND VALVE TESTS). A loss of ALL Off-Site power occurs resulting in the following:

11 EDG tripped on overspeed

Bus 15 is de-energized with no faults

Bus 16 experienced a LOCKOUT condition What procedure must be used to reset the 11 EDG overspeed trip and restore Bus 15?

A. C.4-B.09.06.C (LOSS OF BUS 15 OR BUS 16)

B. C.6-08-B-35 (NO 11 DIESEL ENG MAINTENANCE LOCKOUT)

C. 0187-01 (11 EDG/ESW QUARTERLY PUMP AND VALVE TESTS)

D. B.09.08-05.H.2 (LOCAL SHUTDOWN OF 11 DIESEL IN EMERGENCY CONDITION)

CORRECT ANSWER:

C JUSTIFICATION: 0187-01 (11 EDG/ESW QUARTERLY PUMP AND VALVE TESTS) states that if offsite power is lost while the EDG is loaded and in parallel with off-site power and the EDG trips on overspeed, then reset the overspeed trip, place speed droop knob to zero, and depress and release both engine stop pushbuttons. These actions will restart the EDG and provide power to Bus 15.

A is incorrect: This AOP would be entered but does not provide steps to restore the EDG.

B is incorrect: This ARP would be received and entered for an overspeed trip but does not provide steps to restore the EDG.

D is incorrect: This procedure would be used to shutdown the tripped EDG, but since the plant is currently in a SBO a restart would be required. Also plausible as the engine stop pushbuttons need to be depressed to restart the EDG.

SRO Only: The CRS would brief the procedures continuous awareness step in the PJB and direct the actions if the event occurred. Not RO responsibility.

REFERENCE:

0187-01 B.09.08-05 10 CFR:

55.43b(5)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None SRO ONLY JUSTIFICATION:

Assessment of facility conditions and selection of appropriate procedure.

QUESTION SOURCE:

Bank - 2020 NRC Exam - Previous two exams TIER:

2 GROUP:

1 CATEGORY:

264000 EDGs K/A:

A2.09 IMPORTANCE:

SRO 4.3 COG LEVEL:

2 DR K/A DESCRIPTION: Ability to (a) predict the impacts of the following on the EDGs; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of a safety bus.

DIFFICULTY:

3 LESSON PL:

M8107L-042 OBJECTIVE:

9

2022 MONTICELLO ILT NRC EXAM - KEY

91.

The plant was in MODE 1 when an RBCCW line break required the #11 Recirc Pump to be tripped. Actions have been taken to stabilize the plant and the following conditions exist:

Reactor power is 42%

A jet pump loop flow is 7 Mlb/hr

B jet pump loop flow is 32 Mlb/hr Which Technical Specification actions below, if any, are correct for the above conditions?

(C.2-06 Figure 1 Power/Flow Map is on following page)

A. Be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to satisfy 2/3 core height reflood analyses.

B. Match Recirc loop jet pump flows within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to satisfy LOCA analyses.

C. Ensure the limits of LCO 3.2.1, 3.2.2 and 3.3.1.1 are applied when applicable.

D. No TS action is required since the plant remains outside Stability Regions I and II.

CORRECT ANSWER:

C JUSTIFICATION: In single loop operation TS 3.4.1 allows single loop operation provided the plant is not in the extended flow window domain and provided the limits of LCO 3.2.1, 3.2.2 and 3.3.1.1 are applied. The plant is not in the extended flow window.

A is incorrect: Plausible action requirement and reason for a failed jet pump.

B is incorrect: Plausible and correct if the examinee believes 11 recirc pump is still running but its only the reverse flow in the loop.

D is incorrect: Plausible misapplication of the core flows (adding vs. subtracting) adding would make flow 39 Mlb/hr and could allow for vertical misinterpretation of the EFW boundary of the P/F map.

REFERENCE:

TS 3.4.1 and Bases 10 CFR 55.43b(2)

REFERENCE PROVIDED DURING EXAM:

TS 3.4.1, TS 3.4.2 and embedded PF Map SRO ONLY JUSTIFICATION TS Action determination and knowledge of bases.

QUESTION SOURCE:

SM - 2013 NRC Exam #93 TIER:

2 GROUP:

2 CATEGORY:

202001 Recirculation K/A:

A2.22 IMPORTANCE:

SRO 3.2 COG LEVEL:

3SPR/SPK K/A DESCRIPTION:

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: Loss of component cooling water.

DIFFICULTY 2

LESSON PL:

M8107L-029 OBJECTIVE:

10

2022 MONTICELLO ILT NRC EXAM - KEY

92.

The plant is in MODE 3. Given the following:

BOTH airlock doors for the DW personnel airlock have seal damage

At 1530 on July 4th, BOTH airlock doors are declared inoperable

Total containment leakage, including personnel airlock leakage, is 0.55 La Which of the following describes the Tech Spec ACTIONS that are required to be taken?

A. Primary containment is declared inoperable The ACTIONS of TS 3.6.1.1 and TS 3.6.1.2 are entered B. Either one of the two doors is closed by 1630 on July 4th The door that is closed is LOCKED by 1530 on July 5th C. The inner door ONLY must be verified closed by 1630 on July 4th The inner door is restored to OPERABLE status by 1530 on July 5th D. Either one of the two doors closed by 1630 on July 4th Restore the airlock to OPERABLE status by 1530 on July 5th CORRECT ANSWER:

D JUSTIFICATION: In Mode 3, with both airlock doors inoperable, one door must be verified closed within an hour and the airlock must be restored to operable status within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

A is incorrect: The overall containment leakage is < the acceptable limit of 1.0 La. TS 3.6.1.1 would not be entered.

B is incorrect: The closed door is not required to be locked.

C is incorrect: Only restoring the inner door to OPERABLE status is a plausible conservative action, but not required..

REFERENCE:

TS 3.6.1.2 TS 3.6.1.1 TS 5.5.11 10 CFR:

55.43b(2)

EXTERNAL REFERENCE PROVIDED DURING EXAM:

TS 3.6.1.1, TS 3.6.1.2 and TS 5.5.11 SRO ONLY JUSTIFICATION:

Tech Spec action determination QUESTION SOURCE:

Bank - Edits to stem and choices TIER:

2 GROUP:

2 CATEGORY:

290001 Secondary Containment K/A:

A2.01 IMPORTANCE:

SRO 3.3 COG LEVEL:

3 SPR K/A DESCRIPTION: Ability to (a) predict the impacts of the following on the Secondary Containment and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operations: Personnel airlock failure.

DIFFICULTY:

3 LESSON PL:

M8107L-044 OBJECTIVE:

10

2022 MONTICELLO ILT NRC EXAM - KEY

93.

At 0600, the plant is at normal rated conditions. ALL RCS Operational LEAKAGE rates have been stable for 3 days.

The previous 24-hour LEAKAGE calculations are as follows:

Average total LEAKAGE is 24.5 gpm

Average unidentified LEAKAGE is 3.5 gpm At 0700, average total LEAKAGE rises to 26.2 gpm and the following is indicated:

Which procedure must be entered and what TS ACTION/COMPLETION TIME is correct?

(Assume the increased LEAKAGE remains constant and no operator action is taken.)

A. C.4-B.04.01.F (LEAKS INSIDE PRIMARY CONTAINMENT)

The plant must be placed in MODE 3 by 1900 for Pressure Boundary LEAKAGE.

B. C.4-B.01.04.A (TRIP OF ONE RECIRC PUMP)

The plant must be placed in MODE 3 by 1900 for Pressure Boundary LEAKAGE.

C. C.4-B.04.01.F (LEAKS INSIDE PRIMARY CONTAINMENT)

The total LEAKAGE must meet the LCO statement by 1100.

D. C.4-B.01.04.A (TRIP OF ONE RECIRC PUMP)

The total LEAKAGE must meet the LCO statement by 1100.

2022 MONTICELLO ILT NRC EXAM - KEY CORRECT ANSWER:

C JUSTIFICATION: AOP C.4-B.04.01.F (Leaks inside Primary Containment) would be entered and would direct evaluation to TS 3.4.4. Total leakage exceeds LCO 3.4.4.c (<25 gpm). This requires restoration in 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

A is incorrect: Reactor Recirculation pump seal leakage is not pressure boundary leakage.

B is incorrect: The pump would not be immediately trip so the trip of a recirc pump procedure would not be entered.

Reactor Recirculation pump seal leakage is not pressure boundary leakage.

D is incorrect: The pump would not be immediately trip so the trip of a recirc pump procedure would not be entered.

REFERENCE:

C.4-B.04.01.F LCO 3.4.4 & Bases 10 CFR:

55.43b(2)

REFERENCE PROVIDED DURING EXAM:

TS 3.4.4 SRO ONLY JUSTIFICATION:

Facility operating limitations in the technical specifications and their bases.

QUESTION SOURCE:

Bank - 2015 NRC Exam #92 TIER:

2 GROUP:

2 CATEGORY:

290002 Rx Vessel Internals K/A:

2.4.4 IMPORTANCE

SRO 4.7 COG LEVEL:

3 SPK/SPR K/A DESCRIPTION: Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

DIFFICULTY:

3 LESSON PL:

M-8107L-101 OBJECTIVE:

10.b

2022 MONTICELLO ILT NRC EXAM - KEY

94.

The plant is at rated conditions. You are the CRS and a load reduction is about to commence to remove a Reactor Feed Pump from service.

Which of the following is the minimum required reactivity adjustment oversight required for the performance of this load reduction?

A. Duty Control Room Supervisor ONLY B. Duty Control Room Supervisor and Shift Manager ONLY C. Duty Control Room Supervisor, Shift Manager and Nuclear Engineer D. Duty Control Room Supervisor, Shift Manager and Reactivity Management SRO (RM SRO)

CORRECT ANSWER:

D JUSTIFICATION: The Duty shift supervision must verify that the required minimum staffing is met for reactivity adjustments.

FP: Reactivity Manager (RM) - Active Senior Reactor Operator responsible for direct oversight of the manipulation of reactivity controls. During significant reactivity manipulations (R1), the RM SRO shall have no other concurrent duties.

Significant Reactivity Manipulation (R1): Example - Reactor power load changes [PWR]10% [BWR] 15% or greater, not including flexible power operations.

OWI: A member of the Operations Management Team (Corporate Operations to Operations Manager direct reports) should provide management oversight during planned reactivity manipulations, other than Flexible Power Operations.

Examples of these manipulations include:

Load drops and restorations of 25% RTP or less that include Control Rod adjustments that are larger in nature (as determined by Operations Manager). Load drops and restorations with major balance of plant manipulations (e.g.

starting or stopping a Reactor Feed or Recirculation Pump). For removal of the RFP, the examinee must recognize that power will have to be reduced to 50% RTP.

A is incorrect: This would be correct for planned reactivity manipulation of short duration and with a low probability of mis-operation. For example: Recirc flow adjustments to maintain rated power, control rod exercise and stall flow testing.

B is incorrect: This would be correct for non-routine manipulations of longer duration with an elevated probability of mis-operation. For example: Load drops up to 25% power for rod pattern adjustments or MSIV testing.

C is incorrect: The Nuclear Engineer would not fulfill this position

REFERENCE:

OWI-01.06 FP-OP-COO-21 10 CFR 55.43b(6)

REFERENCE PROVIDED DURING EXAM:

None SRO ONLY JUSTIFICATION Knowledge of SRO control room staffing responsibilities for reactivity management.

QUESTION SOURCE:

Bank TIER:

3 GROUP:

1 CATEGORY:

Conduct of Operations K/A:

2.1.5 IMPORTANCE

SRO 3.9 COG LEVEL:

2 DR K/A DESCRIPTION: Ability to use procedures related to shift staffing, such as minimum crew compliment or overtime limitations.

DIFFICULTY 2

LESSON PL:

M8108L-038 OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

95.

The plant is in MODE 5 with a core offload in progress. Complete the following statement:

The basis of the Control Rod OPERABILITY - Refueling Technical Specification is to A. ensure ALL control rods remain fully inserted.

B. prevent AND mitigate prompt reactivity excursion events.

C. ensure ALL control rod scram accumulators are OPERABLE.

D. prevent movement of ANY control rod while in Refueling Mode.

CORRECT ANSWER:

B JUSTIFICATION: During a refueling outage, control rod operability can come into question during various Mode 5 evolutions. The Applicable safety analyses of TS 3.9.5 are the prevention and mitigation of prompt reactivity excursions during refueling.

A is incorrect: One control rod may be withdrawn in Mode 5.

C is incorrect: Scram accumulators are covered in a different LCO.

D is incorrect: Movement of one control rod may be performed.

REFERENCE:

TS 3.9.5 Bases 10 CFR:

55.43b(2)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

None SRO ONLY JUSTIFICATION:

Knowledge of TS Bases QUESTION SOURCE:

Bank - 2018 NRC Exam #94 - Previous 2 exams TIER:

3 GROUP:

1 CATEGORY:

Conduct of Operations K/A:

2.1.36 IMPORTANCE:

SRO 4.1 COG LEVEL:

1B K/A DESCRIPTION: Knowledge of procedures and limitations involved in core alterations.

DIFFICULTY:

3 LESSON PL:

M8107L-019 OBJECTIVE:

10

2022 MONTICELLO ILT NRC EXAM - KEY

96.

The plant is in MODE 2 performing a reactor startup.

Given the following at 1000:

Control Rod withdrawal commenced

Average SRM count rate was 20 cps Given the following at 1300:

The highest achieved average SRM count rate was 160 cps

The reactor was then scrammed and all rods fully inserted If control rod withdrawal is scheduled to recommence at 1600; which of the following is correct in regards to initiating Form 2150 (PRESTART CHECKLIST) and why?

A. A NEW checklist must be initiated because the reactor would have been critical.

B. The ORIGINAL checklist is still valid because < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months elapsed with control rods withdrawn C. The ORIGINAL checklist is still valid because < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months have elapsed since control rods were fully inserted.

D. A NEW checklist must be initiated because > 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months have elapsed since control rod withdrawal commenced.

CORRECT ANSWER:

C JUSTIFICATION: For the given conditions, the examinee must determine that the reactor would not have gone critical (3 doublings). Additionally, they must recognize the correct time requirement for the need to initiate a new 2150 which is

< 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from an all-rods-in condition to re-commencement of rod withdrawal. Neither of these conditions is met so the original 2150 is still valid.

A is incorrect: A new one doesnt need to be initiated. The reactor wouldnt be critical after 3 doublings.

B is incorrect: The original is still valid but not because rods were out for < 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

D is incorrect: A new one doesnt need to be initiated.

REFERENCE:

C.1 Form 2150 Form 2167 10 CFR:

55.43b(1)

EXTERNAL REFERENCE PROVIDED DURING EXAM:

None SRO ONLY JUSTIFICATION:

Initiation and performance of Form 2167 is a SRO ONLY task.

QUESTION SOURCE:

Bank - 2010 NRC Exam #96 TIER:

3 GROUP:

2 CATEGORY:

Equipment Control K/A:

2.2.7 IMPORTANCE

SRO 3.6 COG LEVEL:

3 SPK K/A DESCRIPTION: Knowledge of the process for conducting infrequently performed tests or evolutions.

DIFFICULTY:

2 LESSON PL:

M8113L-001 OBJECTIVE:

5

2022 MONTICELLO ILT NRC EXAM - KEY

97.

Which of the following activities REQUIRE the use of the Bypass Control Process?

A. LIFTING an electrical lead for annunciator 5-B-45 (DRYWELL HI/LO PRESS).

B. Installing a TEMPORARY hose to drain the Service Water System to the floor drains.

C. Opening C-31 INSTALLED Knife Switch 16 with plant power fed from the 1R Transformer.

D. Installing new Condensate Demineralizers that have NOT yet been RELEASED to the plant.

CORRECT ANSWER:

A JUSTIFICATION: AWI-04.04.03 BYPASS CONTROL is the process used for this and states When alarm circuitry is modified directly (annunciator card removal, lifted lead, jumper or shutting off computer alarm points), a Bypass Tag or TMod Tag SHALL be used to identify the circuit modification and a Temporary Information Tag SHALL be placed on the alarm stating it is disabled.

B is incorrect: Does NOT apply to Temporary hoses connected to system drains and routed to floor drains.

C is incorrect: Does NOT apply to Permanently installed switches (test, key, thumbwheel, knife, joystick, handswitch, etc.

D is incorrect: Does NOT apply to Equipment under construction and NOT released to the plant.

REFERENCE:

AWI-04.04.03 Sec 2.1 10 CFR 55.43b(3, 5)

REFERENCE PROVIDED DURING EXAM:

None SRO ONLY JUSTIFICATION SRO knowledge and responsibilities of work control process.

QUESTION SOURCE:

Bank - 2015 NRC Exam #97 - Edits to correct answer TIER:

3 GROUP:

2 CATEGORY:

Equipment Control K/A:

2.2.5 IMPORTANCE

SRO 3.2 COG LEVEL:

1P K/A DESCRIPTION: Knowledge of the process for making design or operating changes to the facility, such as 10 CFR 50.59, Changes, Tests and Experiments, screening and evaluation processes, administrative processes for temporary modifications, disabling annunciators, or installation of temporary equipment.

DIFFICULTY 3

LESSON PL:

M8108L-039 OBJECTIVE:

2

2022 MONTICELLO ILT NRC EXAM - KEY

98.

The plant was at rated conditions when a HPCI steam supply line break occurred. All attempts to isolate the leak have NOT been successful. Conditions are as follows:

RM A-14, HPCI Turbine Area, indicates 1R/hr

TR-4926, Point 26, 896 HPCI Turbine Area, indicates 215ºF and rising

TR-4926, Points 27 and 28, 896 HPCI Turbine Area, indicate 210ºF and rising

Annunciator 4-A-11 (REACTOR BUILDING HI RADIATION) is in alarm

Annunciator 3-B-56 (HIGH AREA TEMPERATURE STEAM LEAK) is in alarm Based on plant conditions, what action is the CRS required to direct?

A. Insert a manual Scram ONLY B. Commence a normal plant shutdown C. Insert a manual Scram AND initiate a Blowdown D. Insert a manual Scram AND if Point 27 or 28 exceed 212ºF then initiate a Blowdown.

CORRECT ANSWER:

A JUSTIFICATION: With the conditions above, SBGT will be running resulting in an elevated release through the plant stack. EOP-1300 directs that if a primary system is discharging into the reactor building and cannot be isolated and any area temperature, radiation, or water level reaches max safe valve (Table T) then scram. This will mitigate the severity of any future release. This criteria is met with a HPCI room temperature >212ºF or rad level of 1R.

B is incorrect: The leak cannot be isolated; therefore, the override in step 26 applies directing entry into step 27.

C is incorrect: There is only 1 area of each parameter that exceeds max safe.

D is incorrect: These points exceeding the max safe are still in the same area; therefore, there is still only 1 area of each parameter that exceeds max safe.

REFERENCE:

C.5-1300 C.5.1-1001 10 CFR:

55.43b(4)

REFERENCE PROVIDED DURING EXAM:

C.5-1300/1400 with entry conditions blanked out SRO ONLY JUSTIFICATION:

EOP action determination and direction QUESTION SOURCE:

Bank - 2015 NRC Exam TIER:

3 GROUP:

2.3 CATEGORY

Radiation Control K/A:

2.3.11 IMPORTANCE:

SRO 4.3 COG LEVEL:

3SPR K/A DESCRIPTION: Ability to control radiation releases.

DIFFICULTY:

3 LESSON PL:

MT-ILT-EOP-004L OBJECTIVE:

4.a and 5.a

2022 MONTICELLO ILT NRC EXAM - KEY

99.

An emergency has occurred with the following conditions:

Off-site Monitoring Teams are projected to receive a thyroid dose of 13 rem CDE

Personnel in security buildings are projected to receive a thyroid dose of 27 rem CDE

A.2-304 (POTASSIUM IODIDE USE) is provided As the Emergency Director, which one of the following describes the course of action regarding Potassium Iodide use?

Issue KI to...

A. personnel in security buildings ONLY but NOT to pregnant and nursing women.

B. personnel in security buildings ONLY but NOT to people with known iodine allergies.

C. off-site Monitoring Teams AND personnel in security buildings but NOT to pregnant and nursing women.

D. off-site Monitoring Teams AND personnel in security buildings but NOT to people with known iodine allergies.

CORRECT ANSWER:

B JUSTIFICATION: The FDA's Bureau of Radiological Health and Drugs Bulletin Volume XVI, Number 7, recommends the issuance of KI to individuals projected to receive a thyroid dose of 25 Rem CDE or more. 25 rem CDE Thyroid is 1000 DAC-hrs and is equivalent to an intake of ~25 uCi of I-131. If the projected thyroid dose, to affected personnel, has exceeded or is projected to exceed 25 Rem CDE, then the only people who should not take potassium iodide are people who know they are allergic to iodine. One may take potassium iodide even if you are taking medicines for a thyroid problem. (For example, a thyroid hormone or anti-thyroid drug.) Pregnant and nursing women may also take this drug.

A is incorrect: Pregnant women may take iodine.

C is incorrect: Off-site monitoring teams need not take iodine because their dose will not exceed 25 Rem and pregnant and nursing women may take iodine.

D is incorrect: Off-site monitoring teams need not take iodine because their dose will not exceed 25 Rem.

REFERENCE:

A.2-304 10 CFR:

55.43b(4)

ADDITIONAL REFERENCE PROVIDED DURING EXAM:

A.2-304 SRO ONLY JUSTIFICATION:

Emergency Director duties.

QUESTION SOURCE:

Bank - 2020 Audit Exam TIER:

3 GROUP:

4 CATEGORY:

Emergency Procedures / Plan K/A:

2.4.40 IMPORTANCE:

SRO 4.5 COG LEVEL:

1P K/A DESCRIPTION: Knowledge of SRO responsibilities in emergency plan implementing procedures.

DIFFICULTY:

2 LESSON PL:

MT-BEP-OPS-001L OBJECTIVE:

9

2022 MONTICELLO ILT NRC EXAM - KEY 100.

The plant was at rated conditions when an LONOP and LOCA occurred. HPCI and RCIC have been unable to maintain RPV water level and the following conditions are now present:

15 Bus LOCKOUT has occurred

Drywell pressure is 8 psig and rising slowly

RPV pressure is 550 psig and lowering slowly

RPV water level is -110 inches and lowering slowly

12 and 14 RHR pumps are running in the Torus Spray/Cooling Mode

12 Core Spray pump is running with MO-1752 (#12 CS OUTBOARD ISOLATION)

OPEN and MO-1754 (#12 CS INBOARD ISOLATION) CLOSED As the CRS, which one of the following directions must be given at this time?

A. Initiate an Emergency Depressurization.

B. Start an additional low pressure ECCS pump.

C. Secure Torus Sprays and realign B RHR to LPCI Mode.

D. Place the 12 Core Spray System in its normal injection lineup.

CORRECT ANSWER:

A JUSTIFICATION: With the recent revision to the MNGP EOPs, Emergency Depressurization should be directed prior to reaching -150 when it is determined that water level cannot be maintained > -126 (TAF). Currently, high pressure systems cannot maintain level and RFPs are not available with the LONOP. ED should be performed.

Distracter B is incorrect. With the given conditions, Division 1 ECCS pumps would be unavailable due to the 15 bus lockout and all Division 2 low pressure ECCS pumps are already running.

Distracter C is incorrect. With the current lineup there is only one injection source lined up with a pump running (12 CS).

An ED should be performed prior to realigning 12 and 14 RHR pumps for injection as these will auto align to LPCI.

Distracter D is incorrect. This is the normal line up for Core Spray.

REFERENCE:

C.5-1100 C.5.1-1001 Bases 10 CFR:

55.43b(5)

REFERENCE PROVIDED DURING EXAM:

EOP 1100 Flowchart with entry conditions blanked out.

SRO ONLY JUSTIFICATION:

EOP Flowchart navigation, action determination and knowledge of bases.

QUESTION SOURCE:

Significantly Modified - 2015 NRC Exam #99 TIER:

3 GROUP:

4 CATEGORY:

Emergency Procedures / Plan K/A:

2.4.23 IMPORTANCE:

SRO 4.4 COG LEVEL:

3 SPK/SPR K/A DESCRIPTION: Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

DIFFICULTY:

3 LESSON PL:

MT-ILT-EOP-002L OBJECTIVE:

4

ML23012A168

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