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Fer 2020 Ile Administered-Written Exam and Handouts
	ML22158A215
Person / Time
Site:	Fermi
Issue date:	01/14/2021
From:	Randy Baker; NRC/RGN-III/DRS/OLB
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	Baker R
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ML19128A206	List: ML22158A208; ML22158A212; ML22158A215;
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CORRECTED ANSWER KEY By RDB at 6:14 pm, Feb 03, 2021 D

&B

1 K/A Importance: 3.5/3.6 Points: 1.00 R01- Difficulty: 3.00 Level of Knowledge: Source: NEW 90269 POST Fundamental SUBMIT VERSIO N

The plant was at 100% power when a Main Turbine Trip and ATWS occurred.

Following completion of initial ATWS actions, conditions are as follows:

SLC is injecting.

Reactor power is 22%.

RPV water level is 75 and steady.

Several SRVs are lifting.

The CRS now directs the CRLNO to lower RPV water level.

Which of the following is a valid reason for lowering RPV water level further?

RPV water level would be lowered to reduce...

A. natural circulation and reactor power.

B. the fraction of voids in the core region.

C. steam flow to the value of Minimum Core Steam Flow (MCSF).

D. downcomer water level below the elevation of the feedwater spargers.

Answer: A ILT 2020 Final Version Page: 1 of 259 Question 1 Approved View

Answer Explanation:

Conditions in the stem of the question indicate that RPV water level would be below Level 2, the point where both recirculation pumps would be tripped. The candidate must recall that, under these conditions, as RPV water level decreases, the height of the fluid columns is reduced, thereby reducing the natural circulation driving head through the core. This lowers the natural circulation flow which raises the void fraction, thereby adding negative reactivity and lowering reactor power.

Distractor Explanation:

B is incorrect because lowering RPV level will lower natural circulation which will raise the amount of voids in the core causing void fraction to rise. Plausible if applicants incorrectly recall the mechanism of subcooling and void formation in the core.

C is incorrect because the level reduction for the conditions presented in the stem of the question is not related to Minimum Core Steam Flow (MCSF). This distractor is plausible because another set of conditions further down the ATWS RPV Level Control leg do require terminating and preventing injection to lower RPV injection rate to (as close as practicable) to the Minimum Core Steam Flow Injection Rate (MCSFIR), which would lower steam flow to near MCSF. However, this reduction does not occur until RPV water level is <0", which is not true as presented in the stem of the question.

D is incorrect because, when RPV level is lowered below 114 as part of initial ATWS actions, the feedwater spargers will have already been uncovered. Plausible because this is the reason for lowering RPV level during initial ATWS actions.

Reference Information:

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

295001 AK1.01 Natural circulation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

Core and Fuel Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: Identify abnormal and emergency operating procedures associated with the Core and Fuel.

ILT 2020 Final Version Page: 2 of 259 Question 1 Approved View

2 K/A Importance: 4.1/4.2 Points: 1.00 R02- Difficulty: 3.00 Level of Knowledge: Source: BANK: 2018 NRC 97228 POST Fundamental EXAM.

SUBMIT TAL VERSIO N

If an EDG was already running in parallel with Offsite Power, how will the EDG respond to a loss of Offsite Power?

The EDG output breaker will...

A. trip and the EDG will shut down and remain shut down.

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

C. trip, the EDG will shut down, and then it will automatically restart in isochronous mode.

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

Answer: B ILT 2020 Final Version Page: 3 of 259 Question 2 Approved View

Answer Explanation:

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

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Although the output breaker will trip, the EDG will not shut down. This distractor is plausible if the candidate thought the EDG would trip and require manual operator action to restart, which is a common failure inserted into simulator scenarios requiring candidates to manually restart an EDG.

C. Although the EDG output breaker will trip, the EDG itself will remain running therefore the EDG will be in isochronous but does not shut down and restart.

D. The EDG output breaker does not remain closed because the breaker trips open due to the function of the under frequency relay as described above. The other parts of this distractor are correct because the EDG does remaining running and its governor will shift to isochronous.

Reference Information:

ST-OP-315-0065 EDG Student Text i-n-2711-36 EDG 14 Control Drawing Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 295003 Partial or Complete Loss of A.C. Power 295003 AK2 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C.

POWER and the following:

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

Fermi 2 NRC Exam Usage ILO 2018 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental RO Associated objective(s):

ILT 2020 Final Version Page: 4 of 259 Question 2 Approved View

3 K/A Importance: 2.6/3.1 Points: 1.00 R03 Difficulty: 2.00 Level of Knowledge: Source: NEW 89867 Fundamental Following a station blackout, AOP 20.300.SBO, Loss of Offsite and Onsite Power, directs operators to reduce BOP and ESF DC electrical loads if AC power cannot be restored to battery chargers after A. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to ensure sufficient DC power is available to start the Blackstart Unit, CTG-11-1, manually.

B. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to ensure sufficient DC power is available to start the Blackstart Unit, CTG-11-1, manually.

C. 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to maintain the electrical and instrumentation components needed for Core Cooling and Decay Heat Removal.

D. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months to maintain the electrical and instrumentation components needed for Core Cooling and Decay Heat Removal.

Answer: C ILT 2020 Final Version Page: 5 of 259 Question 3 Approved View

Answer Explanation:

20.300.SBO condition AC, steps AC.1 and AC.2 direct that attachments 11 and 12 be performed if battery chargers cannot be restored within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . These attachments shed DC loads on BOP and ESF busses respectively.

20.300.SBO BASES states:

Action AC.1-AC.2 This action reduces loading on respective batteries. Failure of the Blackstart Unit to start or if BOP busses cannot be restored requires shutdown of BOP DC powered equipment when plant conditions permit. This condition is outside the assumptions analyzed in the UFSAR. The Blackstart Unit is designated as an alternate AC power source for the plant and is available within one (1) hour to the blacked out unit. Plant coping is controlled predominately by Class IE DC power and steam driven sources until the alternate AC power is available for loading. UFSAR 8.4.2.1 states the SBO coping duration calculated for Fermi 2 is four (4) hours. The specific SBO duration is based on the redundancy of the onsite emergency AC power sources, the reliability of the onsite emergency AC power sources, the expected frequency of loss of offsite power, and the probable time needed to restore offsite power. To maintain the electrical and instrumentation components needed for Core Cooling and Decay Heat Removal following SBO, Fermi 2 requires Class IE as well as non-IE batteries to support operation of the alternate AC source. A battery capacity calculation has been performed to verify that required Class IE and non-IE batteries have sufficient capacity to meet Station Blackout loads for one hour.

Distractor Explanation:

A. Is incorrect because DC power for Blackstart Unit CTG 11-1 is not a factor or basis for performing load reduction of BOP and ESF DC busses. Although CTG 11-1 is essential to restore AC power to the site, it has its own DC supply needed for starting the unit and is not affected by BOP or ESF DC loads. Applicants may choose this if they incorrectly recall the DC supply needed for starting CTG 11-1.

B. Is incorrect because load shedding is directed after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , not 4; and DC power for Blackstart Unit CTG 11-1 is not a factor or basis for performing load reduction of BOP and ESF DC busses. Although CTG 11-1 is essential to restore AC power to the site, it has its own DC supply needed for starting the unit and is not affected by BOP or ESF DC loads. Applicants may choose this if they incorrectly assume that DC load shed timing is based on the SBO coping time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months noted in the UFSAR; and incorrectly recall the DC supply needed for starting CTG 11-1.

D. Is incorrect because load shedding is directed after 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , not 4. Applicants may choose this if they incorrectly assume that DC load shed timing is based on the SBO coping time of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months noted in the UFSAR.

K/A Reasoning:

This question meets the k/a since applicants must recall the reasons for performing DC load shedding following a loss of AC power to battery chargers.

Reference Information:

20.300.SBO, Loss of Offsite and Onsite Power, and applicable bases ILT 2020 Final Version Page: 6 of 259 Question 3 Approved View

Plant Procedures 20.300.SBO Loss of Offsite and Onsite Power NUREG 1123 KA Catalog Rev. 2 295004 Partial or Complete Loss of D.C. Power 295004 AK3. Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER :

295004 AK3.01 2.6/3.1 Load shedding: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.

10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

DC Electrical Distribution Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Discuss design considerations, capabilities, and limitations related to DC Electrical Distribution System component operation.

ILT 2020 Final Version Page: 7 of 259 Question 3 Approved View

4 K/A Importance: 3.6/3.6 Points: 1.00 R04 Difficulty: 3.00 Level of Knowledge: Higher Source: NEW 89887 cognitive level The plant is operating with the following conditions:

Reactor power is 26% and stable.

Main Turbine 1st stage pressure is 158 psig and stable.

Main Generator protective relaying trips the Main Generator due to an electrical fault.

The result of this is A. The Main Turbine trips and the reactor remains at power.

B. The Main Turbine trips and a FULL reactor scram occurs.

C. Main Turbine control valves will rapidly close due to the load rejection and a FULL reactor scram will occur.

D. Main Turbine control valves will throttle steam flow to reduce turbine speed and the reactor will remain at power.

Answer: A ILT 2020 Final Version Page: 8 of 259 Question 4 Approved View

Answer Explanation:

When the protective relays trip the main generator a main turbine trip signal is initiated, closing TSVs and TCVs.

With reactor power below 30% and main turbine 1st stage pressure below the setpoint of 161.9 psig, the scram signals generated by TSV closure and TCV fast closure will be bypassed. Therefore, no scram signal will be generated in this instance.

Distractor Explanation:

B. Is incorrect because the TSV closure (turbine trip) scram signal will be bypassed in this instance due to turbine 1st stage pressure below the bypass setpoint. Applicants may select this if they incorrectly recall the turbine trip scram bypass signal setpoint and assume a scram will occur.

C. Is incorrect because the TCV fast closure (load reject) scram signal will be bypassed in this instance due to turbine 1st stage pressure below the bypass setpoint. Applicants may select this if they incorrectly believe the TCV fast closure scram will be active under these conditions.

D. Is incorrect because the generator protective relay trip will cause the main turbine to also trip.

Applicants may select this if they incorrectly believe the turbine will remain un-tripped with the TCV fast closure scram bypassed due to turbine 1st stage pressure.

K/A Reasoning:

This question meets the selected k/a since applicants must predict (and be able to monitor) the response of the RPS system to a turbine trip signal under given plant conditions.

Reference Information:

ARP 3D91, TURBINE STOP/CONT VAL CHANNEL TRIP BY-PASSED 23.109, Turbine Operating Procedure 23.601, Instrument Trip Sheets Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 295005 Main Turbine Trip 295005 AA1. Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP:

295005 AA1.02 3.6/3.6 RPS 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO Associated objective(s):

Main Generator and Excitation Cognitive Terminal In accordance with approved plant procedures/references, given various controls and indications for operation of the Main Generator and Excitation System: Discuss the system interrelationships with other plant systems.

ILT 2020 Final Version Page: 9 of 259 Question 4 Approved View

5 K/A Importance: 4.1/4.1 Points: 1.00 R05 Difficulty: 2.00 Level of Knowledge: Source: NEW 90328 Fundamental The plant is operating at 100% power when the Inboard MSIV for one main steam line fast closes.

After the plant stabilizes, the following indications are observed:

Reactor Power is 106%.

RPV Level is 194.

RPV Pressure is 1096 psig.

Which of the following is required?

A. Place the MODE Switch in Shutdown.

B. Lower the Pressure Regulator Setpoint.

C. Place the Feedwater DCS switch in Single Element.

D. Lower Power to <60% using Reactor Recirculation Flow.

Answer: A ILT 2020 Final Version Page: 10 of 259 Question 5 Approved View

Answer Explanation:

The candidate must evaluate plant conditions and interpret that Reactor Pressure is above the High RPV Scram Setpoint of 1093 psig (see ARP 3D75). With the Reactor still at power, the candidate must interpret that a failure to scram has occurred and the correct action is the place the Mode Switch in Shutdown.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. Reactor Pressure is above the normal value and the candidate could fail to recognize that pressure is above the scram setpoint. With pressure above 1045 psig, Tech Specs requires pressure be restored below 1045 psig within 15 minutes, so the candidate could determine this is the correct course of action. However, the High RPV Scram Setpoint has been exceeded, requiring a reactor scram.

C. Because the MSIV closed, there exists an indicated steam flow/feedwater flow mismatch, which is causing RPV Level to be controlling off of the normal setpoint of 197. The candidate could recognize this and determine that placing DCS in Single Element is required to correct the deviation, which would be true if the High RPV Scram Setpoint had not been exceeded, requiring a reactor scram. However, the High RPV Scram Setpoint has been exceeded, requiring a reactor scram.

D. The candidate could recall that the procedural guidance in 23.137 for closing one MSIV at power requires lowering power <60%, which is how the crew would normally proceed for isolating one main steam line at power. If the candidate failed to recognize the high RPV pressure setpoint being exceeded, this action would lower pressure, and power, to within normal values and establish conditions for one main steam line isolated. However, the High RPV Scram Setpoint has been exceeded, requiring a reactor scram.

Reference Information:

3D75, Reactor Vessel High Pressure Channel Trip.

23.137 Nuclear Boiler SOP Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 295006 SCRAM 295006 AA2. Ability to determine and/or interpret the following as they apply to SCRAM :

295006 AA2.04 4.1/4.1* Reactor pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

Nuclear Boiler System Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: Discuss the Nuclear Boiler system interrelationships with other systems.

ILT 2020 Final Version Page: 11 of 259 Question 5 Approved View

ILT 2020 Final Version Page: 12 of 259 Question 5 Approved View

6 K/A Importance: 4.4/4.7 Points: 1.00 R06 Difficulty: 4.00 Level of Knowledge: Higher Source: BANK 90608 cognitive level A startup was in progress when a fire occurred in the Turbine Building.

Turbine Building Area Temperature exceeded 200°F and an automatic Reactor Scram occurred.

All control rods are fully inserted and the Reactor MODE Switch is in SHUTDOWN.

Conditions require the evacuation of the Main Control Room to the Remote Shutdown Panel.

The Relay room is accessible.

At the Remote Shutdown Panel:

RPV Pressure is 900 psig.

RPV Water Level is 175 inches.

If COLD Shutdown is desired, which of the following is (1) The procedurally directed method for conducting a cooldown?

(2) The LOWEST RPV Pressure allowed, by the procedure, within ONE hour?

A. (1) LOWER the pressure setting on pressure controllers in Turbine Control Relay Panel.

(2) 400 psig.

B. (1) LOWER the pressure setting on pressure controllers in Turbine Control Relay Panel.

(2) 350 psig.

C. (1) OPEN Safety Relief Valves A or B at the Remote Shutdown Panel.

(2) 400 psig.

D. (1) OPEN Safety Relief Valves A or B at the Remote Shutdown Panel.

(2) 350 psig.

Answer: C ILT 2020 Final Version Page: 13 of 259 Question 6 Approved View

Answer Explanation:

NOTE: 20.000.19 Attachment 1, page 1, Saturated Steam Tables to be provided with this question.

RISING Turbine Building Area Temperatures causing a Reactor Scram indicates MSIVs are CLOSED.

SRVs will be operated. 20.000.19 limits cooldown to 90°F/hr. The LOWEST RPV Pressure allowable within ONE hour is 400 psig.

Distracter Explanation:

(1) LOWER the pressure setting on pressure controllers in Turbine Control Relay Panel is plausible and incorrect if the examinee thought the cause of the scram was not closed MSIVs and thought they were OPEN.

(2) 350°F is plausible and incorrect if the examinee incorrectly believed the allowable cooldown limit is 100°F/hr.

Reference Information:

AOP 20.000.19 Cond K pg 13 and Attachment 1, pg 1 This question matches the selected K/A because RO applicants must utilize reactor pressure instrument readings and make an operational decision regarding cooldown method and rate following control room abandonment.

Plant Procedures 20.000.19 Question Use ILO Reference Provided RO NUREG 1123 KA Catalog Rev. 2 G2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Higher Cognitive Level Reference Provided RO ILT 2020 Final Version Page: 14 of 259 Question 6 Approved View

Associated objective(s):

Remote Shutdown System Cognitive Terminal In accordance with approved plant procedures/references, given various controls and indications for operation of the Remote Shutdown System: Discuss effective monitoring and control of the Remote Shutdown System using local and remote controls, indications, computer displays, alarms, and data-logging devices.

ILT 2020 Final Version Page: 15 of 259 Question 6 Approved View

7 K/A Importance: 3.5 Points: 1.00 R07 Difficulty: 4.00 Level of Knowledge: High Source: BANK 91687 The plant is operating at 100% power with the following:

Center Station Air Compressor is RUNNING.

West Station Air Compressor is in AUTO.

How would the Station Air System respond if all TBCCW flow is lost to the Station Air Compressors?

A. The Center compressor will trip immediately. The East and West compressors will not start either automatically or manually.

B. The Center compressor will continue to run until it trips on high temperature.

The East and West compressors will not start, either automatically or manually.

C. The Center compressor will trip immediately. The West compressor will not automatically start. The East can be manually started but will trip on high temperature.

D. The Center compressor will continue to run until it trips on high temperature.

The West compressor will auto start and run until it trips on high temperature.

The East can be manually started but will trip on high temperature Answer: C ILT 2020 Final Version Page: 16 of 259 Question 7 Approved View

Answer Explanation:

Note: This is a BANK question from the Fermi 2 exam bank that has NOT been previously used on any NRC exam.

Per 23.129 Enclosure A and 7D61, the Center and West Station Air Compressors have trips on Low (cooling) Water Flow. This will cause the Center compressor to trip on loss of TBCCW and the West compressor will not automatically start.

Also per the same references above, the East compressor does not have a trip associated with loss of TBCCW.

Therefore, the candidate should conclude that the Center compressor will trip immediately (on low cooling water flow), the standby (West) compressor will not automatically start, however the East compressor can be started but it will eventually trip on high temperature.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could assume that all three compressors have trips on low cooling water flow, which would prevent both the West and East compressors from starting. This is incorrect because the East compressor does not have a trip associated with loss of TBCCW so it can be started but will eventually trip on high temperature due to no cooling water flow.

B. The candidate could assume that low cooling water flow is only a permissive to start and not a trip once the compressor is running, which could lead the candidate to determine that the Center will remain running, with no cooling water flow, until it trips on high temperature but the East and West will not start (either in automatic or manually) due to the no cooling water flow permissive. This is incorrect because the Center compressor will trip immediately (on low cooling water flow), the standby (West) compressor will not automatically start, and the East compressor can be started but it will eventually trip on high temperature.

D. The candidate could assume that all 3 compressors are like the East and therefore determine that all 3 will run, with no cooling water flow, until they trip on high temperature. This is correct for the East but incorrect for the Center and West since they have trips on low cooling water flow.

Reference Information:

23.129, Station Air System, Enclosure A, SAC Alarm/Trip Setpoints.

7D61 SAC Trouble.

NUREG 1123 KA Catalog Rev. 2 295018 Partial or Complete Loss of Component Cooling Water 295018 AK1.01 3.5/3.6 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Effects on component/system operations 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (4) Secondary coolant and auxiliary systems that affect the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO ILT 2020 Final Version Page: 17 of 259 Question 7 Approved View

Associated objective(s):

Turbine Building Closed Cooling Water Cognitive Terminal In accordance with approved plant procedures/references, given various controls and indications for operation of the TBCCW System: Discuss the system interrelationships with other plant systems.

ILT 2020 Final Version Page: 18 of 259 Question 7 Approved View

8 K/A Importance: 3.3 Points: 1.00 R08- Difficulty: 3.00 Level of Knowledge: High Source: NEW 97229 POST SUBMIT TAL VERSIO N

Division 2 Non-Interruptible Air System (NIAS) has been lost.

What is the impact of this loss on the ability of the T2300-F450B, Torus to RB Vacuum Breaker, and the T2300-F410, Torus to RB Vacuum Breaker Iso Valve, to perform their design functions?

The vacuum relief function is __(1)__.

The containment isolation function is __(2)__.

A. (1) lost (2) lost B. (1) maintained (2) lost C. (1) lost (2) maintained D. (1) maintained (2) maintained Answer: D ILT 2020 Final Version Page: 19 of 259 Question 8 Approved View

Answer Explanation: D Per 20.129.01, Loss of Station and/or Control Air AOP, Enclosure A (System Response on Loss of Div 2 NIAS) for the Torus Vacuum Breakers:

T2300-F450B, Torus to RB Vacuum Breaker, will be closed on loss of air because air is used to test the vacuum breaker in the open direction only. The vacuum breaker will still actuate to perform its design function to limit excessive primary containment negative pressure. Additionally, with this check valve in its normally closed condition on loss of air, the containment isolation function is maintained.

T2300-F410, Torus to RB Vacuum Breaker Iso Valve, fails open on a loss of air, allowing the design function to limit excessive primary containment negative pressure to be maintained.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This combination of distractors is plausible if the candidate incorrectly recalls that NIAS is needed to (1) operate the T2300-F450B, Torus to RB Vacuum Breaker in the open direction to prevent excessive primary containment negative pressure and (2) to close the T2300-F410. Part (1) is incorrect because NIAS is not needed to actuate the vacuum breaker, however, this is plausible because NIAS is supplied to the vacuum breaker, but only to stroke the valve open for testing, not for it to perform its design function of relieving excessive negative containment pressure. Part (2) is correct because air is needed to close T2300-F410 and it fails open on loss of Div 2 NIAS.

B. This combination of distractors is plausible if the candidate incorrectly recalls that the T2300-F410, Torus to RB Vacuum Breaker Iso Valve fails closed on loss of Div 2 NIAS. If this were true, the containment function would be maintained, however, the ability of the Torus Vacuum Breakers to limit excessive primary containment negative pressure would be lost. This distractor is incorrect because the T2300-F410 and it fails open on loss of Div 2 NIAS.

C. This combination of distractors is plausible if the candidate incorrectly recalls that the air supply to the T2300-F410, Torus to RB Vacuum Breaker Iso Valve, was another air source, such as Div 1 NIAS, and therefore is unaffected by the loss of Div 2 NIAS. This is plausible because the majority of containment penetrations at Fermi 2 are supplied with 2 isolation valves powered (either electrically or via air) from opposite sources. The candidate could incorrectly recall that the T2300-F410, Torus to RB Vacuum Breaker Iso Valve is powered from Div 1 NIAS since the T2300-F450B, Torus to RB Vacuum Breaker is a Division 2 valve. This would allow both the containment isolation and vacuum relief functions to be maintained. This is incorrect because T2300-F410, Torus to RB Vacuum Breaker Iso Valve, fails open on loss of Div 2 NIAS and not Div 1 or any other air source.

Reference Information:

20.129.01, Loss of Station and/or Control Air AOP.

NUREG 1123 KA Catalog Rev. 2 295019 Partial or Complete Loss of Instrument Air 295019 AK2 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR and the following:

295019 AK2.09 3.3/3.3 Containment 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO ILT 2020 Final Version Page: 20 of 259 Question 8 Approved View

Associated objective(s):

ILT 2020 Final Version Page: 21 of 259 Question 8 Approved View

9 K/A Importance: 3.3/3.4 Points: 1.00 R09 Difficulty: 2.00 Level of Knowledge: Higher Source: NEW 90628 cognitive level The plant is in MODE 4.

Shutdown Cooling is in service using Div 1 RHR pump A.

RHRSW Pumps A & C are in service

RHR Div 2 is unavailable Which of the following would require the crew to implement 23.800.05, Alternate Reactor Coolant Circulation and Decay Heat Removal Core Spray or RHR, to perform a feed and bleed?

A. C7102-S001A, RPS Motor Generator Set A trips.

B. E1151-C001C, RHRSW pump C trips on overcurrent.

C. E1102-C002A, RHR pump A, experiences a shaft shear.

D. E1150-F008, RHR SDC Otbd Iso Vlv, fails shut and cannot be opened.

Answer: D ILT 2020 Final Version Page: 22 of 259 Question 9 Approved View

Answer Explanation:

AOP 20.205.01, Loss of Shutdown Cooling, directs operators to initiate 23.800.05 when SDC cannot be re-established in either loop. This procedure establishes feed & bleed conditions using RHR or Core Spray, and SRVs.

The only condition that would necessitate feed and bleed is the failure of F008, since this valve is required to be open for SDC to function in either loop.

A is incorrect since AOP 20.205.01 provides actions to restore RPS, reset isolations, and restore SDC using the loop previously in service. It is plausible if applicants incorrectly determine that SDC cannot be restored using following a loss of RPS A power.

B is incorrect since RHRSW pump A is still available to provide RHRSW flow to the RHR heat exchanger.

It is plausible if applicants incorrectly determine that SDC cannot be restored due to the loss of one RHRSW pump.

C is incorrect since RHR pump C is available to provide SDC flow. It is plausible if applicants incorrectly determine that SDC cannot be restored, or that a feed and bleed lineup is required.

References:

20.205.01, Loss of Shutdown Cooling 23.800.05, Alternate Reactor Coolant Circulation and Decay Heat Removal Core Spray or RHR This question matches the selected K/A because RO applicants must evaluate plant conditions and determine which variation of a loss of SDC will necessitate performance of a feed and bleed.

Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 295021 Loss of Shutdown Cooling 295021 AK3. Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING :

295021 AK3.02 3.3/3.4 Feeding and bleeding reactor vessel 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (3) Mechanical components and design features of reactor primary system.

10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New NRC Early Review RO Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Performance Enabler Recognize and respond to a Loss of Shutdown Cooling ILT 2020 Final Version Page: 23 of 259 Question 9 Approved View

10 K/A Importance: 3.3/3.5 Points: 1.00 R10 Difficulty: 4.00 Level of Knowledge: Higher Source: MODIFIED: 2018 AND 90347 cognitive level 2019 RETAKE EXAMS A refueling accident has caused the following:

16D1, RB Refueling Area Fifth Floor High Radiation is received.

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

3D35, Div I/II FP Vent Exh Radn Monitor Upscale Trip is received.

All Fuel Pool Vent Exhaust Radiation Monitors peaked at 5.5 mR/hr and are now slowly lowering.

(1) What automatic actions, if any, will occur AND (2) which of the following procedure(s) entry conditions is (are) met?

A. (1) No automatic actions occur.

(2) 20.710.01, Refueling Floor High Radiation, AND 20.000.02, Abnormal Release of Radioactive Material ONLY B. (1) RBHVAC isolates.

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

(2) 20.710.01, Refueling Floor High Radiation, AND 20.000.02, Abnormal Release of Radioactive Material ONLY D. (1) RBHVAC isolates.

(2) 20.710.01, Refueling Floor High Radiation, 20.000.02, Abnormal Release of Radioactive Material AND 29.100.01, Sheet 5 Secondary Containment Control Answer: D ILT 2020 Final Version Page: 24 of 259 Question 10 Approved View

Answer Explanation:

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

Per 3D35 if the setpoint of 3mr/hr is reached, RBHVAC will isolate, CCHVAC will shift to the Recirculation Mode, and SGTS will automatically start. Also, if FP Vent Exh Rad Monitors are verified greater than 3 mr/hr, perform 20.000.02, Abnormal Release of Radioactive Material and 20.710.01, Refueling Floor High Radiation. Finally, if Fuel Pool Vent Exh Duct Radiation is verified greater than 5 mr/hr, perform 29.100.01 SH 5, "Secondary Containment and Rad Release".

Therefore, the candidate must recall the above and determine that (1) RBHVAC WILL isolate and (2) 20.000.02, 20.710.01 AND 29.100.01, Sheet 5 must be entered.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Note: This was the previously correct answer on the 2018 exam version. This distractor is plausible because the examinee could have memorized the bank version but failed to recognize that, in this version, FP Vent Exh Rad Monitors have exceeded their trip setpoints, which will cause RBHVAC to isolate.

B. Part (1) is correct in that RBHVAC isolates. Part (2) is plausible if the candidate recognizes the need to enter Abnormal Release of Radioactive Material due to the RB5 radiation levels but failed to recognize that entry conditions also exist for 20.710.01 and 29.100.01, Sheet 5.

C. Note: This was the previously correct answer on the 2019 retake exam version. This distractor is plausible because the examinee could have memorized the bank version but failed to recognize that, in this version, FP Vent Exh Rad Monitors have exceeded 5 mr/hr thus requiring 29.100.01, Sheet 5 to be entered. Part (1) is correct in that RBHVAC isolates. Part (2) is incorrect because 29.100.01 is also required to be entered if FP Vent Exh Rad levels exceed 5 mr/hr.

Reference Information:

20.710.01, Refueling Floor High Radiation.

16D1, RB Refueling Area Fifth Floor High Radiation 3D35, Div I/II FP Vent Exh Radn Monitor Upscale Trip.

ILT 2020 Final Version Page: 25 of 259 Question 10 Approved View

Plant Procedures 16D01 Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 295023 Refueling Accidents 295023 AA1. Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS 295023 AA1.01 3.3/3.5 Secondary containment ventilation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

10 CFR 55.41(b)(13) Procedures and equipment available for handling and disposal of radioactive materials and effluents.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level Modified RO Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Performance Enabler Recognize, respond to, and correct fuel cladding failure (small)

ILT 2020 Final Version Page: 26 of 259 Question 10 Approved View

11 K/A Importance: 3.9/3.9 Points: 1.00 R11 Difficulty: 3.00 Level of Knowledge: Higher Source: NEW 90427 cognitive level A plant transient has resulted in the following conditions:

Drywell pressure.........9.2 psig

Torus pressure............5.7 psig

Drywell Temperature.....235°F

Torus temperature.......105°F All parameters above are rising slowly.

The crew should initiate Drywell sprays...

A. after Torus pressure exceeds 9.0 psig.

B. after Torus temperature exceeds 110°F.

C. now, because Drywell pressure has exceeded 9.0 psig.

D. now, because Drywell temperature cannot be maintained below 242°F.

Answer: A ILT 2020 Final Version Page: 27 of 259 Question 11 Approved View

Answer Explanation:

Per 29.100.01 sheet 2, PC Control, drywell sprays are initiated under 2 conditions. Either torus pressure exceeding 9.0 psig, or drywell temperature exceeds 242°F and before exceeding 340°F. In this instance, torus pressure is not yet at the value requiring DW sprays, nor is DW temp. Therefore, the crew must wait until either torus pressure reaches 9# or it is determined that DW temp cannot be maintained below 340°F.

Distractor Explanation:

B. Is incorrect since torus temperature exceeding 110° is not a driver for DW sprays. It is plausible because 110° torus temp is a milestone that requires other EOP actions.

C. Is incorrect since no parameter requires DW sprays now. Further, DW sprays are initiated based on torus pressure, NOT DW pressure. It is plausible because this is a common point of confusion in the performance of the PC control EOP WRT initiation of DW sprays.

D. Is incorrect since no parameter requires DW sprays now. DW sprays should not be initiated until after 242°F, and other actions are taken first. It is plausible since DW temp is a trigger for DW spray initiation.

K/A Reasoning:

This question matches the selected K/A because RO applicants must interpret the current and future value of torus pressure in a high drywell pressure event and determine the correct crew actions.

Reference Information:

Per 29.100.01 sheet 2, PC Control BWROG EPG/SAG Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 295024 High Drywell Pressure.

295024 EA2. Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

295024 EA2.04 Suppression chamber pressure: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO Associated objective(s):

Primary Containment Control Cognitive Terminal Given appropriate procedures and conditions, Describe the plant conditions that would require use of the alternative actions contained in 29.100.01 Sh 2, Primary Containment Control, including: in accordance with plant/management expectations: a. Emergency Depressurization; b.

Torus Spray; c. Drywell Spray; d. Venting the Drywell; ILT 2020 Final Version Page: 28 of 259 Question 11 Approved View

ILT 2020 Final Version Page: 29 of 259 Question 11 Approved View

12 K/A Importance: 3.8/4.5 Points: 1.00 R12 Difficulty: 4.00 Level of Knowledge: Higher Source: NEW 90268 cognitive level The plant is operating at 100% power with the #1 pressure regulator out of service.

How will the plant respond, and what procedure(s) will be entered, if the #2 pressure regulator were to fail LOW?

Reactor pressure will...

A. Rise; Enter 20.129.02, Reactor Pressure Controller Failure and lower the Pressure Regulator setpoint.

B. Lower; Enter 20.129.02, Reactor Pressure Controller Failure and verify indicated Pressure Regulator Setpoints are the same as before the failure.

C. Lower; Enter 20.129.02, Reactor Pressure Controller Failure and trip the Main Turbine and Bypass Valves, then enter 29.100.01, Sheet 1 - RPV Control.

D. Rise; Enter 29.100.01, Sheet 1 - RPV Control and place the MODE Switch in Shutdown. Then enter 20.129.02, Reactor Pressure Controller Failure to disable the failed Pressure Regulator.

Answer: D ILT 2020 Final Version Page: 30 of 259 Question 12 Approved View

Answer Explanation:

The candidate must recognize that the conditions presented in the stem of the question will cause Reactor Pressure to rise above the High RPV Pressure Scram Setpoint, which is an entry condition for EOP Per 29.100.01, Sheet 1 - RPV Control. The candidate must then recognize that, to restore the ability to control pressure using the Main Steam Bypass Valves, AOP 20.129.02 must be entered to disable the failed Pressure Regulator (Conditions C and D).

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Reactor Pressure does rise if the in-service Pressure Regulator signal fails low, and this action is what the operator would take if the backup pressure regulator were able to take control, per Condition B of 20.129.02. However, this distractor is incorrect because, with the #1 Pressure Regulator out of service as specified in the stem of the question, there is no backup to take control so Reactor Pressure would continue to rise to the Scram / EOP entry point.

B. Reactor Pressure would lower if the in-service Pressure Regulator signal were to fail high and the candidate could confuse the verbiage (fails high vs fails low) and therefore incorrectly determine how the plant would respond, which is a common misconception due to the ambiguity of the terms signal fails high vs. signal fails low." Also, the candidate could incorrectly determine that Condition A is applicable, which requires verifying indicated Pressure Regulator Setpoints are the same as before the failure. However, the conditions in the stem indicate that the plant will respond by RPV pressure rising, making this distractor incorrect.

C. Reactor Pressure would lower if the in-service Pressure Regulator signal were to fail high and the candidate could confuse the verbiage (fails high vs fails low) and therefore incorrectly determine how the plant would respond, which is a common misconception due to the ambiguity of the terms signal fails high vs. signal fails low." This could lead the candidate to determine that the correct action is the enter 20.109.02 and place the Mode Switch in Shutdown, then Trip the MS Bypass Valves and Main Turbine, and then enter 29.100.01 - RPV Control, which is the correct set of actions for the Pressure Regulator failing high. However, the conditions in the stem indicate that the plant will respond by RPV pressure rising, making this distractor incorrect.

Reference Information:

29.100.01, Sheet 1 - RPV Control.

20.129.02, Reactor Pressure Controller Failure Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 295025 High Reactor Pressure G2.4.8 Knowledge of how abnormal operating procedures are used with EOPs 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO ILT 2020 Final Version Page: 31 of 259 Question 12 Approved View

Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Performance Enabler Recognize, respond to, and correct Pressure Regulator Signal Fails LOW ILT 2020 Final Version Page: 32 of 259 Question 12 Approved View

13 K/A Importance: 3.5/3.8 Points: 1.00 R13- Difficulty: 4.00 Level of Knowledge: Fund Source: NEW 97407 POST SUBMIT TAL VERSIO N

Which of the responses below best describes why Emergency RPV Depressurization is required when Torus Water Temperature exceeds the Heat Capacity Limit (HCL) Curve?

Torus Water Temperature at the HCL is the highest Torus Water Temperature ...

A. which can occur without steam in the suppression chamber airspace.

B. at which pressure suppression capability sufficient to accommodate an RPV breach by core debris can be maintained.

C. at which opening an SRV will not result in exceeding the code allowable stresses in the SRV tail pipe, tail pipe supports, quencher, or quencher supports.

D. from which Emergency RPV depressurization will not raise suppression chamber pressure above the Primary Containment Pressure Limit before the rate of energy transfer from the RPV to the containment is within the capacity of the containment vent.

Answer: D ILT 2020 Final Version Page: 33 of 259 Question 13 Approved View

Answer Explanation:

Per EPG/SAGs, Appendix B Vol I (Introduction) Section 9.0 Variables and Curves. Section 9.7 for the Heat Capacity Temperature Limit (HCTL, or HCL at Fermi) states that the HCL is the highest suppression pool (Torus Water) temperature from which Emergency RPV depressurization will not raise suppression chamber pressure above the Primary Containment Pressure Limit before the rate of energy transfer from the RPV to the containment is within the capacity of the containment vent. Knowledge of this meets the K/A because the operational implication of exceeding the HCL is the inability to condense steam adequately enough to maintain pressure within the limit of the PCPL curve before heat energy into the Torus is within the capacity of the containment vent. The HCL is a function of Torus Water Temperature, Torus Water Level and RPV Pressure.

The candidate must recall that this is the reason why (the operational implication of) high Torus Water Temperature.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. the highest suppression chamber pressure which can occur without steam in the suppression chamber airspace is the bounding limit for the Pressure Suppression Pressure curve at Fermi 2.

Since the HCL is concerned with maintaining pressure suppression capability, and preventing steam in the suppression chamber airspace, the candidate could confuse the operational implication of exceeding the PSP curve with that of the HCL curve and choose this response. This is incorrect because the calculation for the HCL curve does not take this bounding limit into account.

B. the highest suppression chamber water level at which pressure suppression capability sufficient to accommodate an RPV breach by core debris can be maintained is a bounding limit for determining the Maximum Pressure Suppression Primary Containment Water Level (MPSPCWL) at Fermi 2.

Since the HCL is also concerned with maintaining suppression capability, the candidate could confuse the operational implication of water level being above the MPSPCWL with that of the HCL curve and choose this response. This is incorrect because the calculation for the HCL curve does not take this bounding limit into account.

C. the highest suppression chamber water level at which opening an SRV will not result in exceeding the code allowable stresses in the SRV tail pipe, tail pipe supports, quencher, or quencher supports is the bounding limit for the SRV Tail Pipe Level Limit (SRVTPLL) at Fermi 2. Since the HCL is concerned with maintaining the integrity of containment and components within containment, the candidate could confuse the operational implications of exceeding the SRVTPLL curve with that of the HCL and choose this response. This is incorrect because the calculation for the HCL curve does not take this bounding limit into account.

Reference Information:

EPG/SAGs, Appendix B Vol I (Introduction) Section 9.0 Variables and Curves..

ILT 2020 Final Version Page: 34 of 259 Question 13 Approved View

NUREG 1123 KA Catalog Rev. 2 295026 Suppression Pool High Water Temperature 295026 EK1. Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE :

295026 EK1.02 3.5/3.8 Steam condensation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

ILT 2020 Final Version Page: 35 of 259 Question 13 Approved View

14 K/A Importance: 3.7 Points: 1.00 R14 Difficulty: 2.00 Level of Knowledge: High Source: NEW 93047 Why is Drywell Spray prohibited when Drywell Temperature is above the allowable value of the Drywell Spray Initiation Limit (DWSIL) curve?

To prevent...

A. steam from forming within the suppression chamber airspace.

B. rapid, cyclic condensation of steam at the downcomer openings of the Drywell vents.

C. rapidly lowering containment pressure causing a loss of net positive suction head to pumps taking a suction from the Torus.

D. exceeding the capability of the vacuum relief system and challenging the primary containment negative pressure limit.

Answer: D ILT 2020 Final Version Page: 36 of 259 Question 14 Approved View

Answer Explanation:

Per the EPG Appendix B Definition of the Drywell Spray Initiation Limit (DWSIL), unrestricted operation of drywell spray could result in a negative pressure large enough to [deinert the primary containment atmosphere] or challenge the primary containment negative pressure capability. This question focuses on the second part of this definition.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This is a partial definition of the Pressure Suppression Pressure (PSP) and the candidate could recall this definition but incorrectly apply it to the DWSIL curve and determine that spraying within the bounds of the curve prevents steam formation in the suppression chamber airspace, which is plausible since spraying the drywell will cause the spray water to flash to steam, which must be condensed by the water in the suppression chamber when it is forced through the downcomer vents.

This is incorrect, however, because spraying within the bounds of the DWSIL curve is not related to preventing steam formation in the suppression chamber.

B. This is the definition of the phenomenon known as "chugging". The candidate could incorrectly determine that spraying above the DWSIL could cause rapid, cyclic condensation of steam (or chugging) to occur inside the Drywell, OR the candidate could recall the definition of chugging but incorrectly relate it to the basis for the DWSIL curve, which is incorrect because the basis of the DWSIL curve is given above.

C. Caution 8 warns against reducing primary containment pressure and the fact that it may cause exceeding NPSH limits for pumps taking a suction on the Torus. References to Caution 8 are found throughout 29.100.01 Sheet 2, Primary Containment Control, whenever Drywell (and Torus) Sprays are on. The candidate could partially remember this caution and determine that it is applicable to the DWSIL curve. This is incorrect because the basis of the DWSIL curve is given above.

Reference Information:

EPG Appendix B, Chapter 18, Variables and Curves.

NUREG 1123 KA Catalog Rev. 2 295028 High Drywell Temperature 295028 EK2. Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following:

295028 EK2.01 3.7/4.1 Drywell spray: Mark-I&II 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO Associated objective(s):

ILT 2020 Final Version Page: 37 of 259 Question 14 Approved View

15 K/A Importance: 3.8/4.1 Points: 1.00 R15 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED: FERMI 90387 2015 NRC R15 A plant event has damaged the Reactor Building. The Operating Shift is executing the EOPs.

The reactor is shutdown.

RPV Pressure is 920 psig, stable, being controlled by Turbine Bypass Valves.

29.ESP.27, Torus Leak Isolation, is in progress.

Torus Water level is -30", lowering slowly.

RPV emergency depressurization should be performed ___(1)___ because Torus water level

___(2)___.

A. (1) immediately (2) is now low enough that steam discharged from the drywell into the suppression pool may not be condensed.

B. (1) immediately (2) now corresponds to the Minimum Pressure Suppression Primary Containment Water Level. Below this level, the pressure suppression capability of the primary containment may be insufficient to accommodate an RPV breach by core debris.

C. (1) once Torus water level reaches -38" (2) will then be low enough that steam discharged from the drywell into the suppression pool may not be condensed.

D. (1) once Torus water level reaches -38" (2) will then correspond to the Minimum Pressure Suppression Primary Containment Water Level. Below this level, the pressure suppression capability of the primary containment may be insufficient to accommodate an RPV breach by core debris.

Answer: C ILT 2020 Final Version Page: 38 of 259 Question 15 Approved View

Answer Explanation:

From BWROG EPG:

Suppression pool water level must be maintained above the elevation of the Mark I/II downcomer vent openings or least 2 feet above the top of the Mark III horizontal vents to ensure that steam discharged from the drywell into the suppression pool following a primary system break will be adequately condensed. (Results of the Bodega Bay Mark I containment tests indicate 95% steam condensation may be achieved from a vertical downcomer vent that discharges at a level six inches above the suppression pool surface.) If suppression pool water level cannot be maintained above the specified minimum value, steam may not be adequately condensed, and primary containment pressure could exceed allowable limits. Since the RPV may not be kept at pressure when pressure suppression capability is unavailable, Emergency RPV Depressurization is required.

EOP sheet 3, Primary Containment Control, step TWL-6 requires ED when it is determined that Torus level cannot be maintained >-38".

ODE-10, Operations expectations for EOP usage, page 21 states:

c. If Torus level reaches -30 while performing 29.ESP.27, restore Low Pressure Systems that were isolated per the ESP and evaluate the criteria for anticipate ED.

d. Prior to reaching -38, the crew must be prepared to perform Emergency Depressurization.

Therefore, the crew should not immediately ED in this instance, but are required to when T/L reaches -

38".

Distracters Explanation:

"(1) immediately." - This answer is plausible and incorrect. The examinee would choose this answer because it is a recognized milestone in the performance of the PC control EOP, and ODE-10.

However it is not expected or required that ED be performed at this torus level.

"(2) ...correspond(s) to the Minimum Pressure Suppression Primary Containment Water Level. Below this level, the pressure suppression capability of the primary containment may be insufficient to accommodate an RPV breach by core debris." - This answer is plausible and incorrect. The examinee would choose this answer based on remembering the upper limit for TWL. IE the Maximum Pressure Suppression Primary Containment Water Level.

K/A Reasoning:

This question matches the selected K/A since RO applicants must recall the requirements to ED with lowering torus level, and recall the reasons for performing ED.

Reference Information:

BWR EPG Appendix B (pg B-7-49) and (pg B-17-64) 29.100.01 sh 2, Primary Containment Control 29.ESP.27, Torus Leak Isolation ILT 2020 Final Version Page: 39 of 259 Question 15 Approved View

Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 295030 Low Suppression Pool Water Level 295030 EK3. Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL:

295030 EK3.01 3.8/4.1 Emergency depressurization 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level Modified RO Associated objective(s):

Cautions, Curves and Calculations Cognitive Terminal Given a set of plant parameters that meet the entry conditions, discuss the definition of and reason for the shape of the following graphs/curves in accordance with Fermi 2 Emergency Operating Procedures: a. Boron Injection Initiation Temperature Curve; b. CS/RHR/HPCI/RCIC NPSH Limits; c. CS/RHR Vortext limits; d. Heat Capacity Limit; e. RPV Saturation Temperature; f.

SRV Tail Pipe Level Limit; g. Pressure Suppression Pressure; h. Primary Contaimnet Pressure Limit ILT 2020 Final Version Page: 40 of 259 Question 15 Approved View

16 K/A Importance: 3.6/3.7 Points: 1.00 R16 Difficulty: 3.00 Level of Knowledge: Higher Source: NEW 90369 cognitive level A LOCA has occurred and the following indications exist:

RPV Pressure is 900 psig and lowering.

Drywell Pressure is 2.5 psig and rising.

RPV Water Level is 80" and lowering.

All Condenser pumps are tripped.

HPCI and RCIC are isolated.

SBFW is unavailable.

You have been directed to maximize injection to the RPV using the CRD Pumps in accordance with 29.ESP.04, RPV Injection using CRD Pumps.

Which of the following sets of actions will accomplish this directive?

(1) Place the CRD flow controller in MANUAL.

(2) Start the second CRD Pump.

Then...

A. (3) Close the CRD Flow Control Valve.

(4) Close the CRD Pressure Control Valve.

B. (3) Throttle open the CRD Flow Control Valve.

(4) Open the CRD Pressure Control Valve.

C. (3) Throttle open the CRD Flow Control Valve.

(4) Close the CRD Pressure Control Valve.

D. (3) Close the CRD Flow Control Valve.

(4) Open C1100-F034, CRD Charging Water Header Isolation Valve.

Answer: B ILT 2020 Final Version Page: 41 of 259 Question 16 Approved View

Answer Explanation:

Per 29.ESP.04, the CRD Flow Control Valve is throttled open, and the CRD PCV is opened, to minimize headloss in the system and maximize flow to the RPV via the cooling water header.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This combination of actions is plausible because they would close off flow downstream and maximize pressure on the Charging Header, which would increase flow to the RPV if a scram signal was present. However, flow through this path is restricted because C11-F034 is normally throttled, and the charging header has 4 restricting orifices. Additionally, these actions are not in accordance with 29.ESP.04.

C. This combination of actions is plausible because the candidate could incorrectly recall the flowpath through the CRD system and determine that the cooling water header was between the FCV and PCV, thus concluding that opening the FCV while closing the PCV would increase pressure on / flow through the cooling water header. Furthermore, the actions of opening the FCV and closing the PCV are performed in an ATWS event, so a candidate could recall those actions but fail to recall the conditions under which they are performed. However, closing the PCV will block cooling water flow, which is the desired flowpath as specified in 29.ESP.04.

D. This combination of actions is plausible because they would minimize downstream flow, while maintaining cooling water flow to the CRD drive mechanisms (due to the block on the FCV permitting 15 gpm of flow when shut), but increase pressure and flow through the Charging Header, which would increase flow to the RPV if a scram signal was present. However, flow through this path is still restricted because of the 4 restricting orifices in the charging header. Additionally, these actions are not in accordance with 29.ESP.04.

Reference Information:

29.ESP.04, RPV Injection using CRD Pumps.

M-5703-1, CRD System FOS Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 295031 Reactor Low Water Level.

295031 EA1. Ability to operate and/or monitor the following as they apply to REACTOR LOW WATER LEVEL :

295031 EA1.10 Control rod drive 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO ILT 2020 Final Version Page: 42 of 259 Question 16 Approved View

Associated objective(s):

Emergency Support Procedures Cognitive Terminal Given relevant plant conditions, Describe the steps to maximize control rod drive injection into the Reactor Pressure Vessel, in accordance with plant and management expectations and in accordance with Fermi 2 Emergency Operating Procedures.

ILT 2020 Final Version Page: 43 of 259 Question 16 Approved View

17 K/A Importance: 4.0/4.1 Points: 1.00 R17 Difficulty: 3.00 Level of Knowledge: Higher Source: NEW 90587 cognitive level The plant is in an ATWS event with the following conditions:

Reactor Power is 15%.

RPV Water Level is +20".

SRVs are open, adding heat to the Torus.

Which of the following is the LOWEST Torus Water Temperature above which it will be required to Deliberately Lower RPV Water Level by controlling RPV injection?

A. 95°F.

B. 105°F.

C. 110°F.

D. 120°F.

Answer: C ILT 2020 Final Version Page: 44 of 259 Question 17 Approved View

Answer Explanation:

Per 29.100.01 Sheet 1A, RPV Control - ATWS when Reactor Power is >3% and RPV Level is >0" (Top of Active Fuel), Torus Temperature is >110°F, and an SRV is open (or DW Pressure is >1.68 psig), then the correct action is to Deliberately lower level until power is <3%, or RPV Level = 0", or all SRVs remain closed and Drywell Pressure is <1.68 psig per the 4th statement in FSL-OR1.

Therefore, the candidate must evaluate the conditions of Reactor Power, RPV Water Level ,and SRV status in the stem of the question to conclude that the critical parameter value for Torus Water Temperature under these conditions is 110°F, thus requiring the crew to deliberately lower RPV Water level. Note: 110°F is also the value that requires entry into LCO 3.6.2.1 Condition D.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. 95°F is the Tech Spec LCO limit for Torus Water Temperature in Modes 1, 2 and 3, and it is also the EOP entry value for Torus Water Temperature in the Primary Containment Control EOP. The candidate could recall these relationships and determine that this is the value that also requires deliberately lowering level. However, this is incorrect because, for the conditions given in the stem of the question, deliberately lowering level is not required until Torus Water Temperature reaches 110°F.

B. 105°F is the Tech Spec LCO limit for Torus Water Temperature in Modes 1, 2 and 3 when heat is being added to the Suppression Pool. The candidate could recall this value and determine that the SRVs being open in the stem of the question meets the condition of heat being added to the Suppression Pool and thusly determine that this is the value which also requires deliberately lowering level. However, this is incorrect because, for the conditions given in the stem of the question, deliberately lowering level is not required until Torus Water Temperature reaches 110°F.

D. 120°F is the value that requires entering Tech Spec LCO 3.6.2.1 Condition E, which requires depressurizing the RPV to >200 psig within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . The candidate could recall this significant value and determine that this is also the value that requires deliberately lowering level. However, this is incorrect because, for the conditions given in the stem of the question, deliberately lowering level is not required until Torus Water Temperature reaches 110°F.

Reference Information:

29.100.01, Sheet 1A - RPV Control ATWS LCO 3.6.2.1, Suppression Pool Average Temperature.

Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or unknown.

295037 EA2. Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :

295037 EA2.04 4.0*/4.1* Suppression pool temperature 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO ILT 2020 Final Version Page: 45 of 259 Question 17 Approved View

Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Performance Enabler Execute steps of Anticipated Transient Without Scram to control Reactor Pressure Vessel level (FSL)

ILT 2020 Final Version Page: 46 of 259 Question 17 Approved View

18 K/A Importance: 4.0/4.6 Points: 1.00 R18 Difficulty: 3.00 Level of Knowledge: Source: NEW 90647 Fundamental The three principle Fission Product Barriers are __(1)__. Status of these three barriers are monitored closely, in an emergency, because failure of these may result in __(2)__.

A. (1) Fuel cladding, Reactor Coolant system piping, and Primary Containment (2) excessive offsite radioactive release rates B. (1) Reactor Coolant system piping, Primary Containment, and Secondary Containment (2) excessive offsite radioactive release rates C. (1) Fuel cladding, Reactor Coolant system piping, and Primary Containment (2) excessive Reactor Building rad levels and loss of access to systems required for EOP implementation D. (1) Reactor Coolant system piping, Primary Containment, and Secondary Containment (2) excessive Reactor Building rad levels and loss of access to systems required for EOP implementation Answer: A ILT 2020 Final Version Page: 47 of 259 Question 18 Approved View

Answer Explanation:

Per EP-101:

"... The defense in depth design concept precludes the release of highly radioactive fission products to the environment. This concept relies on multiple physical barriers any one of which, if maintained intact, precludes the release of significant amounts of radioactive fission products to the environment. The primary fission product barriers are:

Fuel Clad (FC): The Fuel Clad Barrier consists of the cladding material that contains the fuel pellets.

Reactor Coolant System (RCS): The RCS Barrier is the reactor coolant system pressure boundary and includes the RPV and all reactor coolant system piping up to and including the isolation valves.

Primary Containment (PC): The Primary Containment Barrier includes the drywell, the suppression pool, their respective interconnecting paths, and other connections up to and including the outermost containment isolation valves. Primary Containment Barrier thresholds are used as criteria for escalation of the ECL from Alert to a Site Area Emergency or a General Emergency.

Distractor Explanation:

B. Is incorrect because secondary containment status is not one of the three defense in depth barriers. It is plausible since the loss of secondary containment may result in the release of fission products.

C. Is incorrect because the concern with the three barriers is the potential offsite release of fission products and the resultant rise in offsite dose. It is plausible since the loss of secondary containment may cause the release of fission products that will likely raise RB rad levels, which may restrict access to plant equipment. However, it is not the rationale for monitoring the status of the three fission product barriers.

D. Is incorrect because secondary containment status is not one of the three defense in depth barriers. It is plausible since the loss of secondary containment may result in the release of fission products. Is also incorrect because the concern with the three barriers is the potential offsite release of fission products and the resultant rise in offsite dose. It is plausible since the loss of secondary containment may cause the release of fission products that will likely raise RB rad levels, which may restrict access to plant equipment. However, it is not the rationale for monitoring the status of the three fission product barriers

References:

EP-101, Classification of Emergencies K/A Reasoning:

This question matches the selected K/A because RO applicants must recall the safety functions evaluated for potential excessive offsite release of fission products.

NUREG 1123 KA Catalog Rev. 2 295038 High Off-Site Release Rate G2.4.21 Knowledge of the parameters and logic used to assess the status of safety functions such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO ILT 2020 Final Version Page: 48 of 259 Question 18 Approved View

Associated objective(s):

Core and Fuel Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: Identify abnormal and emergency operating procedures associated with the Core and Fuel.

ILT 2020 Final Version Page: 49 of 259 Question 18 Approved View

19 K/A Importance: 2.9 Points: 1.00 R19 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 91587 Which of the following correctly completes the statement below regarding fighting a fire in Main Unit Transformer 2A or 2B?

The Transformer __(1)__ System will NOT actuate automatically unless the __(2)__.

A. (1) Deluge (2) Transformers are de-energized B. (1) Deluge (2) Breakers CM and CF are open C. (1) Pre-Action Sprinkler (2) Transformers are de-energized D. (1) Pre-Action Sprinkler (2) Breakers CM and CF are open Answer: B ILT 2020 Final Version Page: 50 of 259 Question 19 Approved View

Answer Explanation:

Per 23.501.01 Section 1.1, System

Description:

A Deluge Sprinkler System is basically a pre-action system with open sprinklers instead of fusible sprinklers. The sprinkler piping is dry until the system Dry Valve is opened by an actuated detector or by manual actuation, locally or from the Control Room. Deluge systems are used when large quantities of water are required for suppression. Once the valve is opened water is discharged from all the sprinklers on the piping system.

Per P&L 3.6: Transformer Deluge Systems will not actuate unless Transformer Output Breakers are open OR pushbutton on COP is depressed and any other initiation signal or fault alarm is in.

Per P&L 3.5: During Turbine Runup, Transformers 2A and 2B may be energized with Output Breakers CM and CF open. In this situation, the Deluge System could actuate and discharge onto the energized transformer(s).

Per P&L 3.7: If Transformer Deluge System actuates, the transformers will not automatically trip.

Transformer Output Breakers must be manually opened.

Therefore, the examinee must first recall that the Main Unit Transformers utilize a Deluge System.

Secondly, the examinee must recall that the Deluge system will not actuate unless the Transformer Output Breakers are open.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. (1) This part is correct. (2) This is plausible because it is reasonable to assume that logic for the Deluge system evaluated the status of the Main Unit Transformers to not allow a massive quantity of water to pour onto them if energized. However, the logic looks at the status of the Output Breakers (CM and CF) and not transformer status.

C. (1) Pre-Action Sprinkler is plausible because this type of sprinkler system, like Deluge, is used in areas where freezing may occur and is functionally similar to Deluge, the only difference being Pre-Action Sprinklers have fusible sprinkler heads and Deluge does not. (2) This is plausible because it is reasonable to assume that logic for the Deluge system evaluated the status of the Main Unit Transformers to not allow a massive quantity of water to pour onto them if energized. However, the logic looks at the status of the Output Breakers (CM and CF) and not transformer status.

D. (1) Pre-Action Sprinkler is plausible because this type of sprinkler system, like Deluge, is used in areas where freezing may occur and is functionally similar to Deluge, the only difference being Pre-Action Sprinklers have fusible sprinkler heads and Deluge does not. (2) This part is correct.

Reference Information:

23.501.01, Fire Water Suppression Systems NUREG 1123 KA Catalog Rev. 2 600000 Plant Fire On Site 600000 AK1. Knowledge of the operation applications of the following concepts as they apply to PLANT FIRE ON SITE:

600000 AK1.02 2.9/3.1 Fire Fighting 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New NRC Early Review RO ILT 2020 Final Version Page: 51 of 259 Question 19 Approved View

Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Performance Enabler Recognize and respond to Plant Fires ILT 2020 Final Version Page: 52 of 259 Question 19 Approved View

20 K/A Importance: 3.1 Points: 1.00 R20 Difficulty: 2.00 Level of Knowledge: Fund Source: NEW 91667 The plant is in MODE 5, Refueling, with RHR Pump A running in Shutdown Cooling (SDC).

RHR Pump B is scheduled to be started, later in the shift, to test the pump. This will support swapping RHR SDC divisions scheduled to take place on the next shift.

The SOC/ITC informs the Main Control Room that there is a system-wide grid disturbance in progress.

AOP 20.300.GRID has been entered.

How should the crew proceed?

A. Test RHR Pump B and swap SDC divisions as scheduled.

B. Shutdown RHR Pump A and start a Reactor Recirculation Pump.

C. Maintain RHR Pump A running and postpone testing of RHR Pump B.

D. Continue with testing RHR Pump B as scheduled but keep RHR Pump A running in SDC.

Answer: C ILT 2020 Final Version Page: 53 of 259 Question 20 Approved View

Answer Explanation:

Per 20.300.GRID, Caution 2 informs operators to Refrain from starting any large equipment (other than those needed for safe operation of the plant) that may have a potential of inducing a transient on the grid.

Condition E (Grid Disturbance in Modes 3, 4 and 5) requires that operators Secure any large electrical loads NOT required to support essential plant operations and Minimize starting Safety Related motors.

Therefore, the examinee must determine that testing of RHR Pump B should be postponed and RHR Pump A left running in SDC.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could fail to recall the requirements to not start any large motors and minimize starting safety related motors. These actions are incorrect because they conflict with the requirements of 20.300.GRID.

B. The candidate could misinterpret guidance in the AOP to do things like restore out of service risk significant equipment (Action D.3) or minimize starting safety related motors (E.2) as meaning all safety related motors should be shut down and determine that RHR must be shut down and a RR pump started for Decay Heat Removal. This is not correct because the AOP does not require this, it only requires securing large electrical loads (such as RHR pumps) that are NOT required to support essential plant operations. RHR in SDC is an essential plant operation.

D. The candidate could fail to recall the AOP requirement to not start any large motors and minimize starting safety related motors and determine that it is acceptable to test RHR Pump B. This is incorrect because the requirement to not start large loads (Caution 2) is there to prevent inducing further transients on the grid and starting a large RHR pump would not be allowed.

Reference Information:

20.300.GRID, Grid Disturbance.

NUREG 1123 KA Catalog Rev. 2 700000 Generator Voltage and Electric Grid Disturbances 700000 AK2. Knowledge of the interrelations between GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the following:

700000 AK2.01 3.1/3.2 Motors 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Performance Enabler Recognize and Respond to Grid Instabilities ILT 2020 Final Version Page: 54 of 259 Question 20 Approved View

21 K/A Importance: 3.4/3.4 Points: 1.00 R21 Difficulty: 2.00 Level of Knowledge: Fund Source: BANK: GRAND GULF 90247 2002 NRC EXAM Which one of the following describes the reason for isolating the Main Steam Isolation Valves on a Low Main Condenser Vacuum?

A. Prevent over-pressurizing the Off Gas system piping that could lead to leakage of radiation to the environment.

B. Prevent erosion damage to turbine blading in the Low Pressure Turbine due to steam condensation in the Main Steam Lines.

C. Prevent over-pressurization of low pressure piping on the suction of the Condenser pumps that could result in a steam rupture inside secondary containment.

D. Prevent the addition of steam that would lead to additional condenser pressurization and possible rupture of the diaphragm installed to protect the turbine exhaust hood Answer: D ILT 2020 Final Version Page: 55 of 259 Question 21 Approved View

Answer Explanation:

Note: The BANK source for this question is the 2002 Grand Gulf NRC ILE Exam.

Per Tech Spec 3.3.6.1 bases:

"The Condenser Pressure-High Function is provided to prevent over pressurization of the main condenser in the event of a loss of the main condenser vacuum. Since the integrity of the condenser is an assumption in offsite dose calculations, the Condenser Pressure-High Function is assumed to be OPERABLE and capable of initiating closure of the MSIVs. The closure of the MSIVs is initiated to prevent the addition of steam that would lead to additional condenser pressurization and possible rupture of the diaphragm installed to protect the turbine exhaust hood, thereby preventing a potential radiation leakage path following an accident."

A is incorrect since the basis is associated with turbine exhaust hood protection. Plausible since the off gas system may become over-pressurized, however the MSIV isolation is not associated with preventing this condition.

B is incorrect since the basis is associated with turbine exhaust hood protection. Plausible since low pressure turbine blading damage is possible under low load conditions, however this is not associated with low condenser vacuum or prevented by MSIV closure.

C is incorrect since the basis is associated with turbine exhaust hood protection. Plausible since the condenser pipe suction piping may experience a rise in pressure, but it would not be sufficient to cause a pipe rupture.

Reference:

Fermi TS 3.3.6.1 bases This question matches the selected K/A since applicants must recall the reason for MSIV isolation following a loss of main condenser vacuum NUREG 1123 KA Catalog Rev. 2 295002 Loss of Main Condenser Vacuum 295002 AK3. Knowledge of the reasons for the following responses as they apply to LOSS OF MAIN CONDENSER VACUUM :

295002 AK3.05 3.4/3.4 Main steam isolation valve: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental RO ILT 2020 Final Version Page: 56 of 259 Question 21 Approved View

Associated objective(s):

Nuclear Boiler System Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Discuss design considerations, capabilities, and limitations related to Nuclear Boiler system component operation.

ILT 2020 Final Version Page: 57 of 259 Question 21 Approved View

22 K/A Importance: 3.1 Points: 1.00 R22 Difficulty: 2.00 Level of Knowledge: Fund Source: BANK: DAEC 2011 90607 NRC Following a loss of all RPV water level indication, 29.100.01, sheet 3, RPV Flooding, directs operators to open Safety Relief Valves (SRVs) and flood the RPV using available injection systems.

When the required number of SRVs are opened, how will the Main Steam Isolation Valves (MSIVs) be positioned and why?

A. Open to allow the RPV to be fully depressurized.

B. Closed to avoid excessive water inventory loss from the RPV during flooding.

C. Closed to prevent damage to downstream lines or hangers due to excessive loading.

D. Open to utilize steam line flow indicators to ensure flooded conditions have been achieved.

Answer: C ILT 2020 Final Version Page: 58 of 259 Question 22 Approved View

Answer Explanation:

Note: BANK source for this question is the 2011 DAEC NRC ILE Exam.

EOP sheet 3, RPV Flooding, directs operators to close MSIVs.

Per the BWROG EPG basis document:

If the specified minimum number of SRVs can be opened or a high pressure motor-driven injection source is available, steam lines connected to the RPV should be isolated. Isolating the steam lines avoids:

Damage resulting from cold water coming in contact with hot metal as the steam lines are flooded.

Excessive loading of lines or hangers not designed to accommodate the weight of water.

Flooding of steam-driven equipment.

A is incorrect since MSIVs are closed when SRVs are available to depressurize the RPV. This is plausible because main steam lines are utilized in other EOP strategies to depressurize the RPV and would be used to depressurize, for RPV flooding, if a sufficient number of SRVs did not open.

B is incorrect since inventory loss down the main steam lines is not a concern. This is plausible if applicants do not correctly understand the reasons for closing the MSIVs as described above.

D is incorrect since MSIVs are closed when SRVs are available to depressurize the RPV, and steam line flow indicators are not used for flooding determination, This is plausible if applicants believe the MSIVs should remain open and are not aware that other indications of flooding conditions are readily available.

References:

29.100.01 sheet 3, RPV Flooding BWROG EPG/SAG This question matches the selected K/A since RO applicants must recall the operating strategy for main steam system components during a high RPV water level condition created by RPV flooding.

NUREG 1123 KA Catalog Rev. 2 295008 High Reactor Water Level 295008 AA1. Ability to operate and/or monitor the following as they apply to HIGH REACTOR WATER LEVEL :

295008 AA1.03 3.1/3.1 Main steam system: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental RO Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Performance Enabler Execute steps of Reactor Pressure Vessel Flooding (RF)

ILT 2020 Final Version Page: 59 of 259 Question 22 Approved View

23 K/A Importance: 3.8 Points: 1.00 R23 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 93227 Which of the following is the LOWEST containment pressure listed that would result in LCO 3.6.1.4, Primary Containment Pressure, NOT being met while in MODES 1, 2 or 3?

A. 1.1 psig.

B. 1.6 psig.

C. 1.8 psig.

D. 2.1 psig.

Answer: D ILT 2020 Final Version Page: 60 of 259 Question 23 Approved View

Answer Explanation:

The candidate must recall that the LCO 3.6.1.4 limit for Containment Pressure is 2.0 psig and evaluate the pressures listed to determine that 2.1 psig is the lowest one listed that makes LCO 3.6.1.4 not met.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. 23.406, Primary Containment Inerting and Purge SOP requires maintaining Primary Containment Pressure between 5 to 19 inches of water. This could lead the candidate to conclude that 1.1 psig would not meet the LCO requirement since 1.1 psig converts to approximately 30" wc, which is well above the normal control band. This distractor is incorrect because, although it would be outside of the normal pressure control band, it meets the LCO 3.6.1.4 requirement.

B. The candidate could relate the setpoint of 3D81, Primary Containment High (1.5 psig) with LCO 3.6.1.4 and determine that if the high alarm setpoint is exceeded then the LCO is not met. This is incorrect because, although 3D81 would be in at 1.6 psig, LCO 3.6.1.4 would still be met.

C. The candidate could relate the setpoint of 3D85, Primary Containment High Pressure Channel Trip (1.68 psig) with LCO 3.6.1.4 and determine that, if the setpoint is exceeded, then the LCO is not met.

This is incorrect because, although 3D85 would be in at 1.68 psig, LCO 3.6.1.4 would still be met.

Reference Information:

LCO 3.6.1.4, Primary Containment Pressure.

23.406, Primary Containment Inerting and Purge.

ARP 3D81, Primary Containment Pressure High/Low.

3D85, Primary Containment High Pressure Channel Trip.

NUREG 1123 KA Catalog Rev. 2 295010 High Drywell Pressure 295010 AA2. Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE :

295010 AA2.02 3.8/3.9 Drywell pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

Containment Systems (T2200 & T2300)

Cognitive Enabler Describe the Containment Systems technical specification limiting conditions for operation, their bases, the associated surveillance requirement(s), and their relationship to operability.

ILT 2020 Final Version Page: 61 of 259 Question 23 Approved View

24 K/A Importance: 4.3/4.4 Points: 1.00 R24 Difficulty: 3.00 Level of Knowledge: High Source: BANK 90610 The plant is in MODE 3 with the following:

Reactor Pressure is 55 psig.

Main Steam Isolation Valves (MSIVs) are open.

Division 1 RHR is operating in Shutdown Cooling with cooldown in progress.

Reactor Water Cleanup is in service.

Power to RPS Bus B is subsequently lost due to a trip of RPS MG Set B, and the crew re-energizes RPS B via the alternate power supply.

Which of the following actions are required?

A. Enter SOP 23.137, Nuclear Boiler System, and reopen the MSIVs. Continue cooldown using Division 1 RHR.

B. Enter AOP 20.707.01, Loss of RWCU, and restore RWCU by resetting the isolation and opening the inboard isolation valve.

C. Enter AOP 20.205.01, Loss of Shutdown Cooling, and restore RHR Loop A by resetting the isolation and opening the outboard suction isolation valve.

D. Enter SOP 23.205, RHR System, and restore Shutdown Cooling by starting an RHR Loop B Pump and opening E1150-F015A, Div 1 LPCI Inboard Isolation Valve.

Answer: C ILT 2020 Final Version Page: 62 of 259 Question 24 Approved View

Answer Explanation:

When RPS B is lost, E1150-F015B Div 2 LPCI Inboard Isolation Valve AND E1150-F008, RHR SDC Outboard Suction Isolation Valve, will close. Closure of the E1150-F015B will not impact RHR in SDC since it is on Division 1. Closure of the E1150-F008 will cause a loss of SDC due to no suction path.

When RPS B is on Alternate Power, the isolation may be reset and the system restarted.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. a HALF Isolation condition will exist on MSIVs , resulting in no MSIV closures. Actions are correct for an MSIV closure. This is also incorrect because Division 1 RHR will trip due to loss of a suction path. Since Division 1 RHR is in SDC, and it is RPS B that was lost, the candidate could conclude that RHR would remain in SDC.

B. it tests a misconception on logic power, Inboard Valve does NOT ISOLATE, G3352-F004 and G3352-F220, RWCU Supply Otbd Iso Vlv and RWCU to FW Otbd Cntm Iso Vlv, will isolate when RPS B is lost.

D. it would be true if the isolation could not be reset and SDC injection established, using Div 1 RHR, via the Div 1 LPCI injection valve. However, if the isolation can be reset, then SDC flow can be re-established via Division 1 RHR.

Reference Information:

SOP 23.601 Pg 11 for E1150-F008 AND E1150-F015B ISOLATION GP AOP 20.205.01 Loss of Shutdown cooling actions.

SOP 23.205.01 Shutdown cooling procedure.

SOP 23.316 Restoring RPS Section 7.3.

Plant Procedures 20.205.01 NUREG 1123 KA Catalog Rev. 2 295020 Inadvertent Containment Isolation G2.1.23 Ability to perform specific system and integrated plant procedures during different modes of plant operation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

Residual Heat Removal Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Describe the normal and alternate power supplies to Residual Heat Removal System components.

ILT 2020 Final Version Page: 63 of 259 Question 24 Approved View

25 K/A Importance: 3.7 Points: 1.00 R25 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 91528 While operating in MODE 3 the following alarms are received:

8D46, Div I Reactor Building Pressure High / Low.

17D46, Div II Reactor Building Pressure High / Low.

Your investigation reveals that RB Differential Pressure is currently -0.050 WC.

What is the status of Secondary Containment?

A. NOT required in the current MODE.

B. OPERABLE because Reactor Building Pressure is negative.

C. INOPERABLE because unmonitored exfiltration of fission products cannot be assured under all conditions.

D. UNKNOWN because the status of secondary containment equipment hatches, pressure relief doors and railroad bay access door seals is not specified.

Answer: C ILT 2020 Final Version Page: 64 of 259 Question 25 Approved View

Answer Explanation:

The candidate must evaluate the conditions and determine that the alarms are in because RB Differential Pressure is low (or, because RB Pressure is high, or not as negative as it should be) because it is above the alarm setpoint of -0.125 WC.

The candidate must then recall that TS LCO 3.6.4.1, Secondary Containment, is Applicable in Modes 1, 2, 3 or whenever moving recently irradiated fuel assemblies in Secondary Containment.

Finally, the candidate must determine that this makes Secondary Containment INOPERABLE.

This question is RO knowledge because information to answer the question is also located in SOP 23.426, Reactor Building HVAC, P&L 3.4, which states "If Secondary Containment vacuum can not be maintained at 0.125 inches of vacuum water gauge, comply with Technical Specifications Section 3.6.4.1, Secondary Containment."

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could fail to recall which MODES Secondary Containment is required to be OPERABLE. Mode 3 is plausible because, if Secondary Containment is INOPERABLE in MODES 1, 2 or 3 and it cannot be restored, the plant is only brought to MODE 3 (see Condition C of LCO 3.6.4.1). This could lead the candidate to believe that Secondary Containment is not required to be OPERABLE in MODE 3, which is incorrect as its applicability is Modes 1, 2, 3 or whenever moving recently irradiated fuel assemblies in Secondary Containment.

B. The candidate could fail to recall the value above which Secondary Containment is required to be maintained to satisfy the LCO, which is plausible because the EOP for Secondary Containment only requires EOP entry when RB Pressure is >0 WC. Since EOP entry conditions are closely tied to TS LCO values in most cases, the candidate could relate the EOP entry condition to TS in this instance and determine that, since RB D/P is not positive, then Secondary Containment must be OPERABLE.

This is incorrect because LCO 3.6.4.1 Surveillance Requirement SR 3.6.4.1.1 requires secondary containment vacuum 0.125 WC.

D. Since the value of Secondary Containment pressure is not contained in the LCO statement (above the line) for LCO 3.6.4.1, the candidate could fail to recall that a specific value is required and instead determine that Secondary Containment OPERABILITY is a function of the status of access openings and hatches to ensure that exfiltration from Secondary Containment will not occur. This is plausible because the status of such openings is verified via SR 3.6.4.1.2 and 3.6.4.1.3. This distractor is incorrect because Secondary Containment can be made INOPERABLE, even with both above SRs met, by RB D/P being above the value allowed by SR 3.6.4.1.1.

Reference Information:

8D46, Div I Reactor Building Pressure High / Low.

17D46, Div II Reactor Building Pressure High / Low.

LCO 3.6.4.1, Secondary Containment.

LCO 3.6.4.1 BASES.

23.426, Reactor Building HVAC, P&L 3.4.

ILT 2020 Final Version Page: 65 of 259 Question 25 Approved View

NUREG 1123 KA Catalog Rev. 2 295035 Secondary Containment High Differential Pressure 295035 EK1 Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE :

295035 EK1.02 3.7/4.2 Radiation release 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (9) Shielding, isolation, and containment design features, including access limitations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

Containment Systems Cognitive Terminal In accordance with approved plant procedures/references, given various controls and indications for operation of the Containment: Identify normal, alarm, and setpoint values for significant monitored parameters in the system.

ILT 2020 Final Version Page: 66 of 259 Question 25 Approved View

26 K/A Importance: 2.8 Points: 1.00 R26- Difficulty: 3.00 Level of Knowledge: High Source: NEW 97231 POST SUBMIT TAL VERSIO N

The plant is operating at 97% power with all Reactor Building watertight doors closed.

A seismic event occurs and the following alarms and indications are received:

6D69, SEISMIC SYSTEM EVENT/TROUBLE

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

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

Sump Level is steady at 31.9 inches.

Flood Level is downscale at 5.0 inches.

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

Sump Level is lowering at 44.0 inches.

Flood Level is downscale at 5.0 inches.

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

Both G1101-D076 RB SW Corner Sump Pumps are running.

(1) Which tank level is rising in Radwaste?

(2) Which system should be isolated to stop the leak?

A. (1) Floor Drain Collector Tank.

(2) HPCI.

B. (1) Equipment Drain Collector Tank.

(2) HPCI.

C. (1) Floor Drain Collector Tank.

(2) Division 2 RHR.

D. (1) Equipment Drain Collector Tank.

(2) Division 2 RHR.

Answer: A ILT 2020 Final Version Page: 67 of 259 Question 26 Approved View

Answer Explanation:

This question requires knowledge of the relationship between sumps located in the Reactor Building and tanks located in Radwaste, which is a direct tie to the K/A.

To answer this question, the candidate should recall that the HPCI room does not have its own floor drain sump and instead drains water, via a trough and pipe, to the D076 (SW floor drain) sump. The candidate should recall that Flood Control Valve T4500-F601 exists in the line between the HPCI room trough and the SW quadrant sump (DO76) to prevent cross-flooding of the SW Quadrant and the HPCI Room.

Since the HPCI Room does not have its own dedicated sump, its floor and trench drains are directed to G1101-D076 (i.e., the SW Quad). Note: T4500-F601 is normally open to allow the HPCI drains to flow into D076. In the event of excessive equipment leakage or a line break, the valve will close on a D076 High/High alarm (+45"). This will prevent water from backing up into the drains and flooding the unaffected area.

The candidate should determine that a leak must have developed in the HPCI room, which caused a high level condition in D076. The candidate should recall that G1101-D076 is a Floor Drain sump and therefore causing a level rise in (1) the Floor Drain Collector Tank.

Since D076 is located in the same room as the Div 2 RHR pumps, the candidate must determine that, since Flood Level in the SW corner is steady and flood level in the HPCI room is rising, that the leak is from the HPCI system and not Div 2 RHR. Therefore, the candidate must conclude that the leak is from (2) the HPCI system, which must be isolated.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could incorrectly recall that D076 is an Equipment Drain Sump and therefore determine that (1) the Equipment Drain Collector Tank level would be rising. This is incorrect because D076 is a Floor Drain Sump and since D076 is a Floor Drain Sump, the (1) Floor Drain Collector Tank level would be rising. The candidate could correctly determine that the leak is from the (2) HPCI system and therefore it must be isolated C. The candidate could correctly recall that D076 is a Floor Drain Sump and therefore determine that (1) the Floor Drain Collector Tank level would be rising. The candidate could incorrectly determine that, since Division 2 RHR Pumps are physically located in the same room as D076, then (2) Div 2 RHR must be isolated. This is incorrect because the stem indicates that D076 sump level is lowering. This is because T4500-F601 has automatically closed, which is leading to D076 Sump Level lowering and increasing water level in the HPCI Room as indicated by G11-R655 rising.

D. The candidate could incorrectly recall that D076 is an Equipment Drain Sump and therefore determine that (1) the Equipment Drain Collector Tank level would be rising. This is incorrect because D076 is a Floor Drain Sump and since D076 is a Floor Drain Sump, the (1) Floor Drain Collector Tank level would be rising. The candidate could incorrectly determine that, since Division 2 RHR Pumps are physically located in the same room as D076, then (2) Div 2 RHR must be isolated.

This is incorrect because the stem indicates that D076 sump level is lowering. This is because T4500-F601 has automatically closed, which is leading to D076 Sump Level lowering and increasing water level in the HPCI Room as indicated by G11-R655 rising.

Reference Information:

29.100.01 SH 5, Secondary Containment Control.

2D78, REAC BLDG FLR/EQUIP DRN SUMPS LVL HI-HI/LO-LO 2D105, REAC BLDG CORNER ROOMS/HPCI ROOM FLOOD LEVEL ILT 2020 Final Version Page: 68 of 259 Question 26 Approved View

NUREG 1123 KA Catalog Rev. 2 295036 Secondary Containment High Sump / Area Water Level 295036 EK2. Knowledge of the interrelations between SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL and the following:

295036 EK2.03 Radwaste 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO Associated objective(s):

ILT 2020 Final Version Page: 69 of 259 Question 26 Approved View

27 K/A Importance: 2.8/3.0 Points: 1.00 R27 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 90807 Per 29.100.01 Sheet 2, Primary Containment Control, why are all available Drywell Fans started prior to venting for High Hydrogen Concentration?

A. Mix hydrogen and oxygen to promote recombination during venting.

B. Lower Drywell Temperature to prevent from exceeding temperatures leading to hydrogen combustion.

C. Redistribute hydrogen throughout the Drywell to prevent localized regions of high hydrogen concentration.

D. Reduce Drywell Pressure to prevent exceeding design limits if deflagration occurs when the vent valves are opened.

Answer: C ILT 2020 Final Version Page: 70 of 259 Question 27 Approved View

Answer Explanation:

Per ST-OP-802-3004-001 section for PC/G control (under Primary Containment Control), the basis for the step (PCG-3) directing starting of DW Fans is "operation of the drywell fans serves to redistribute the hydrogen throughout the drywell, thereby diluting localized regions of high hydrogen concentrations."

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly recall that Primary Containment is vented through the Hydrogen Recombiners, the purpose of which is to recombine hydrogen and oxygen inside containment to maintain each below the minimum required for deflagration, when Containment is vented for high hydrogen concentration. This is incorrect because venting is not through the Hydrogen Recombiners, therefore the reason for running the Drywell Fans is not related to mixing hydrogen and oxygen.

B. Drywell Fans are started in step DWT-3 to draw air across the cooling coils to control (lower) Drywell Temperature and the candidate could determine that Drywell Fans are started in the PCG leg to also lower Drywell Temperature to prevent hydrogen combustion. This is incorrect because Drywell Fans are started to distribute hydrogen around the Drywell and not to prevent from reaching a combustion temperature limit.

D. Per EPG Appendix B Page B-7-66, "if large amounts of hydrogen were to accumulate in the primary containment and ignite in the presence of sufficient oxygen, the peak pressure could exceed the structural capability of the drywell." The candidate could determine that opening the containment vent valves will open a path leading from containment to the outside vent stack and the candidate could conclude that doing so could admit oxygen into containment, thus possibly leading to conditions that would support deflagration as described above, which could cause the candidate to conclude that Drywell Fans are started, as a preventative measure, to reduce pressure (by lowering temperature) before this could occur. This is incorrect because Drywell Fans are started to distribute hydrogen around the Drywell and not to prevent from reaching a Drywell Pressure limit.

Reference Information:

29.100.01 Sheet 2, Primary Containment Control PCG steps.

BWROG EPGs/SAGs, Appendix B, section for PC/G control.

Student Text ST-OP-802-3004-001, Primary Containment Control.t Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 500000 High Containment Hydrogen Concentration.

500000 EK3. Knowledge of the reasons for the following responses as they apply to HIGH CONTAINMENT HYDROGEN CONCENTRATIONS:

500000 EK3.02 2.8/3 Operation of drywell recirculating fans 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO ILT 2020 Final Version Page: 71 of 259 Question 27 Approved View

Associated objective(s):

Primary Containment Control Cognitive Terminal Given appropriate procedures and conditions, Describe the plant conditions that would require use of the alternative actions contained in 29.100.01 Sh 2, Primary Containment Control, including: in accordance with plant/management expectations: a. Emergency Depressurization; b.

Torus Spray; c. Drywell Spray; d. Venting the Drywell; ILT 2020 Final Version Page: 72 of 259 Question 27 Approved View

28 K/A Importance: 3.8 Points: 1.00 R28 Difficulty: 2.00 Level of Knowledge: Fund Source: BANK:FERMI 2012 90187 NRC How is water hammer in RHR/LPCI minimized or prevented?

RHR __(1)__ is maintained full of water by Keep Fill supplied from the __(2)__ System.

A. (1) discharge piping ONLY (2) Demin Water Storage B. (1) suction AND discharge piping (2) Demin Water Storage C. (1) discharge piping ONLY (2) Condensate Storage and Transfer System.

D. (1) suction AND discharge piping (2) Condensate Storage and Transfer System.

Answer: C ILT 2020 Final Version Page: 73 of 259 Question 28 Approved View

Answer Explanation:

RHR discharge piping is supplied by CSTS, suction is supplied from torus via gravity.

A. is incorrect because it is not supplied from the Demin system. Plausible because the CSTS is supplied by the Demin system and because Core Spray keep fill is supplied by the Demin Water Storage Tank.

B. is incorrect because the suction is supplied from the Torus. Is also incorrect because it is not supplied from the Demin system. Plausible because the CSTS is supplied by the Demin system and because Core Spray keep fill is supplied by the Demin Water Storage Tank.

D. is incorrect because the suction is supplied from the Torus. Plausible if applicants incorrectly assume the suction and discharge have the same source.

Reference:

23.205, RHR System OP This question is a match to the selected K/A since applicants must recall RHR/LPCI system design features that prevent water hammer.

NUREG 1123 KA Catalog Rev. 2 203000 RHR/LPCI: Injection Mode 203000 K4. Knowledge of RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following:

203000 K4.05 3.2/3.3 Prevention of water hammer 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILO 2012 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental RO Associated objective(s):

Residual Heat Removal Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Identify the Residual Heat Removal System major components and equipment.

ILT 2020 Final Version Page: 74 of 259 Question 28 Approved View

29 K/A Importance: 3.1 / 3.1 Points: 1.00 R29 Difficulty: 3.00 Level of Knowledge: Higher Source: MODIFIED: FERMI 90468 cognitive level 2019 RETAKE R30 The plant is cooling down in MODE 4, with Division 1 of Shutdown Cooling (SDC) in service

SDC Flow is 10,000 gpm

E1150-F003A, Div 1 RHR Hx Outlet Vlv, is full open

E1150-F048A, Div 1 RHR Hx Bypass Vlv, is throttled You have been directed to decrease the cooldown rate.

Per 23.205, RHR System SOP, what is the proper method to lower the cooldown rate in this condition?

A. Throttle closed E1150-F003A ONLY B. Throttle open E1150-F048A OR Throttle closed E1150-F003A as desired.

C. Throttle closed E1150-F003A, then throttle open E1150-F048A if necessary.

D. Throttle open E1150-F048A, then throttle closed E1150-F003A if necessary.

Answer: D ILT 2020 Final Version Page: 75 of 259 Question 29 Approved View

Answer Explanation:

Per 23.205, section 6.7, Maintaining Temperature while in Shutdown Cooling:

4. Maintain desired cooldown rate by throttling E1150-F003A, Div 1 RHR Hx Outlet Vlv.

5. If necessary to increase cooldown rate after E1150-F003A, Div 1 RHR Hx Outlet Vlv, is fully open, throttle closed E1150-F048A, Div 1 RHR Hx Bypass Vlv.

6. If necessary to decrease cooldown rate, throttle open E1150-F048A, Div 1 RHR Hx Bypass Vlv.

7. If necessary to further decrease cooldown rate after E1150-F048A is fully open, throttle closed E1150-F003A, Div 1 RHR Hx Outlet Vlv.

With the conditions in the stem, to decrease cooldown rate while complying with each of these limits and precautions, it is required to first throttle open on E1150-F048A to reduce the flow through the RHR HX, then.

CAUTION A flow rate of 10,000 to 10,700 gpm should be maintained through the SDC loop.

Less than 10,000 gpm may cause temperature stratification in the Reactor Vessel unless the RPV Head is removed and the cavity flooded. More than 10,700 gpm through the RHR Heat Exchanger could cause damage to the Heat Exchanger.

A limit and precaution for operation in SDC mode states the following:

Either E1150-F048A (B), Div 1 (2) RHR Hx Bypass Vlv, or E1150-F003A (B), Div 1 (2) RHR Hx Outlet Vlv, must be full open to prevent stratification in Rx Vessel.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Incorrect, while throttling F003A closed in this condition would reduce CDR, it is not procedurally permitted, since either F003A or F048A must be full open per the P&L noted above. Plausible since this action would reduce CDR, and applicants may not recall the precaution.

B. Incorrect because the proper sequence of positioning F048A and F003A must be followed in order to prevent violation of the P&L noted above. Plausible because throttling either valve as stated would reduce the CDR, and applicants may not recall the P&L.

C. Incorrect because closing F003A first would violate the P&L noted above Plausible because throttling either valve as stated would reduce the CDR, and applicants may not recall the P&L.

Reference Information:

23.205, RHR System SOP.

ILT 2020 Final Version Page: 76 of 259 Question 29 Approved View

NUREG 1123 KA Catalog Rev. 2 205000 RHR Shutdown Cooling Mode 205000 K5. Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) :

205000 K5.02 2.8/2.9 Valve operation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level Modified RO Associated objective(s):

Residual Heat Removal Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Describe general Residual Heat Removal System operation, including component operating sequence, normal operating parameters, and expected system response.

ILT 2020 Final Version Page: 77 of 259 Question 29 Approved View

30 K/A Importance: 3.8 Points: 1.00 R30 Difficulty: 3.00 Level of Knowledge: High Source: NEW 91591 A loss of feedwater has caused HPCI to start and it is currently maintaining RPV Water Level.

RPV level is 202 and rising.

A subsequent failure of Primary Containment Isolation System (PCIS) logic has caused relay K78 to go to its ENERGIZED position.

Relay K78 has the following attributes:

It is normally DE-ENERGIZED.

It ONLY affects the E4150-F079, HPCI Exh Vac Bkr Inbd Iso Valve.

It is supposed to energize on Low Steam Supply Pressure AND High Drywell Pressure.

How will this malfunction affect the HPCI System?

A. HPCI will IMMEDIATELY trip on High Exhaust Pressure.

B. HPCI will NOT be affected IF the E4150-F075, HPCI Exh Vac Bkr Otbd Iso Vlv remains functional.

C. IF HPCI is shutdown, excess water from the suppression pool will be drawn into the turbine exhaust line causing water hammer on restart.

D. 2D65, HPCI Turbine Exh Drn Pot Level High will alarm causing E4100-F053, HPCI Turb Exh Pot Drn Valve to open requiring HPCI to be shutdown.

Answer: C ILT 2020 Final Version Page: 78 of 259 Question 30 Approved View

Answer Explanation:

The candidate must evaluate conditions in the stem and determine that the E4150-F079 will go closed when the relay in the stem goes to its energized state. The candidate must recall that the normal position for this valve is open and, with the valve closed, the system will continue to operate normally. The candidate must recall that, if HPCI is shutdown, the closed valve will prevent the vacuum breakers from performing their function, thus excess water will be drawn into the HPCI turbine exhaust line on shutdown, which could cause water hammer upon system restart.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly recall that exhaust steam normally flows through the line that is isolated by closure of the E4150-F079, thus resulting in high exhaust backpressure and turbine trip.

This is incorrect because this line has check valves that prevent steam flow through the line and only allow flow from the torus airspace to relieve potential vacuum in the HPCI exhaust line.

B. The candidate could determine that the failure will cause the E4150-F079 to fail open, which would be acceptable if the redundant containment isolation valve (E4150-F075) performed its containment isolation function. This is incorrect because the E4150-F079 will fail closed, which will cause the conditions described above.

D. The candidate could determine that closure of the E4150-F079 will cause a reduction in drain flow from the HPCI exhaust line, which would cause the exhaust line drain pot to fill up, thus resulting in the receipt of 2D65, which would require HPCI to be shutdown. This is incorrect because HPCI exhaust line drain flow is not impacted by closure of the E4150-F079.

Reference Information:

I-2221-10, Schematic for E4150-F075 and F079.

I-2225-03, HPCI Logic (partial).

E41-00, HPCI Design Basis Document (DBD).

NUREG 1123 KA Catalog Rev. 2 206000 HPCI System.

206000 K6. Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE COOLANT INJECTION SYSTEM :

206000 K6.10 3.8/4 PCIS: BWR-2,3,4 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO Associated objective(s):

High Pressure Coolant Injection Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: Describe HPCI System automatic features.

ILT 2020 Final Version Page: 79 of 259 Question 30 Approved View

31 K/A Importance: 3.3/3.2 Points: 1.00 R31 Difficulty: 3.00 Level of Knowledge: Higher Source: BANK: FERMI 2015 90689 NRC The plant is in MODE 4.

Division 1 Core Spray is running in the Test mode per 23.203, Core Spray System.

Flow through DIV 1 Core Spray is being controlled by E2150-F015A, Div 1 CS Test Line Iso Valve, at 3500 gpm.

Due to a logic error a LOCA signal is generated from Div 1 CS Logic ONLY.

How will the Core Spray System respond?

A. Div 1 Core Spray will trip. Div 2 Core Spray will line up to inject.

B. Div 1 and 2 Core Spray will line up to inject. The test valve will close.

C. Div 1 Core Spray will remain running in the TEST mode. Div 2 Core Spray will line up to inject.

D. Div 1 Core Spray will remain running in the TEST mode. Div 2 Core Spray will remain in Standby.

Answer: B ILT 2020 Final Version Page: 80 of 259 Question 31 Approved View

Answer Explanation:

A LOCA signal from either CS Logic will auto start both divisions of Core Spray (I-2215-02).

Per 23.203, Section 5.5, Auto Initiation Div 1, "If Div 1 Core Spray is in the Test Mode when an automatic initiation signal is received, E2150-F015A, Div 1 CS Test Line Iso Vlv, will close as the system aligns for injection to the Reactor Vessel."

Distracter Explanation:

A. Is incorrect and plausible if examinee incorrectly assumes that system lineup would cause a trip condition when an injection signal occurs.

C. Is incorrect and plausible if examinee did not know about the auto closure.

D. Is incorrect and plausible if the examinee felt that a LOCA signal from only Div 1 CS Logic does not provide a system start signal.

Reference Information:

23.203 Section 5.5 (pg 22) NOTE I-2215-02 Core Spray Logic Plant Procedures 23.203 NUREG 1123 KA Catalog Rev. 2 209001 Low Pressure Core Spray System.

209001 A1. Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including:

209001 A1.08 System lineup 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Fermi 2 NRC Exam Usage ILO 2015 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

Core Spray System Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: Describe Core Spray System automatic features.

ILT 2020 Final Version Page: 81 of 259 Question 31 Approved View

32 K/A Importance: 3.6 Points: 1.00 R32 Difficulty: 3.00 Level of Knowledge: High Source: BANK: FERMI 2015 90188 NRC The SLC Initiation Keylock Switch, C4100- M004, has been placed in the PMP A RUN position.

The following indications are noted 30 seconds later:

Reactor Pressure is 1000 psig.

C41-R601, SLC Tank Level Indicator, is steady.

SLC Continuity Lights A and B are ON.

SLC Pump A CMC Switch red light is ON, and green light is OFF.

C41-R600, SLC Pump Discharge Pressure Indicator, is oscillating between 1320 and 1370 psig.

These are indications of (1) what condition, and (2) what action should the operator perform?

A. (1) SLC Explosive Valves failed to fire.

(2) Start SLC Pump B.

B. (1) Normal operation for the SLC System.

(2) Monitor SLC Tank level.

C. (1) C41-F001, SLC Storage Tank Outlet Valve is closed.

(2) Dispatch an operator to open C41-F001.

D. (1) C41-F029A, SLC Pump A Discharge Relief Valve failed open.

(2) Dispatch an operator to gag shut C41-F029A.

Answer: A ILT 2020 Final Version Page: 82 of 259 Question 32 Approved View

Answer Explanation:

If the C41-F004A & B, failed to fire, positive displacement SLC Pump A will OPEN C41-F029A, SLC Pump A Discharge Relief Valve, which causes pressure oscillations between 1320 and 1370 psig.

These are the lift and reseat pressures for C41-F029A. Starting SLC Pump B will fire the other primer in both valves.

Distracter Explanation:

B. Is incorrect and plausible, normal Indication would be discharge pressure slightly higher than Reactor Pressure AND lowering SLC Tank Level.

C. Is incorrect and plausible, the Tank Level would remain steady if the Storage Tank Outlet were shut, but discharge pressure would be low.

D. Is incorrect and plausible, Relief Valve has opened, but has not failed.

Reference Information:

SOP 23.139 (Pg 11&12) SLC injection Plant Procedures 23.139 NUREG 1123 KA Catalog Rev. 2 211000 SLC System 211000 A2. Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

211000 A2.02 3.6/3.9* Failure of explosive valve to fire 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Fermi 2 NRC Exam Usage ILO 2015 Exam ILT 2020 Exam LOR 2019 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

Standby Liquid Control Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Describe general Standby Liquid Control System operation, including component operating sequence, normal operating parameters, and expected system response.

ILT 2020 Final Version Page: 83 of 259 Question 32 Approved View

33 K/A Importance: 3.9/3.8 Points: 1.00 R33 Difficulty: 3.00 Level of Knowledge: Higher Source: BANK 90768 The plant is at 100% power.

A blown fuse causes the Pilot Scram Valve Solenoid light for Trip System A Group 3 Power ON light on H11-P603 to EXTINGUISH.

All other Pilot Scram Valve Solenoid lights on P603 remain LIT Subsequently, 480VAC bus 72E de-energizes.

Which one of the following is the DIRECT result of this event?

A. Annunciator 3D74, Trip Actuators B1/B2 Tripped is received, and reactor power will drop below 50%.

B. Annunciator 3D74, Trip Actuators B1/B2 Tripped is received, and the plant remains at 100% power.

C. Annunciator 3D73, Trip Actuators A1/A2 Tripped is received, and the plant remains at 100% power.

D. Annunciator 3D73, Trip Actuators A1/A2 Tripped is received, and reactor power will drop below 50%.

Answer: A ILT 2020 Final Version Page: 84 of 259 Question 33 Approved View

Answer Explanation:

Each of the 8 pilot scram solenoid lights on P603 indicate power available to the scram solenoids for

~25% of control rod HCUs in each trip system. With one group light off, power has been lost to approximately ~25% of scram solenoids in that trip system.

A loss of bus 72E will result in a loss of power to RPS B trip system, and all trip system B pilot scram solenoids will de-energize. Since ~25% of control rods had a trip system A solenoid de-energized, those control rods will each fully insert and reactor power will drop well below 50%.

B is incorrect because ~25% of all control rods will scram and power will drop significantly. This is plausible if applicants do not recall the impact of a blown fuse for one pilot valve group when a half scram on the opposite trip system.

C is incorrect because the loss of 72E will not affect A RPS, and a 1/4 scram will occur reducing power well below 50%. This is plausible if applicants do not correctly recall the normal power supplies to A & B RPS busses.

D is incorrect because the loss of 72E will not affect A RPS.

References:

23.316, RPS 120V AC AND RPS MG SETS 23.610, Reactor Protection System ARP 3D73, Trip Actuators A1/A2 Tripped ARP 3D74, Trip Actuators B1/B2 Tripped I-2156-01, RPS diagram This question matches the selected K/A because RO applicants must recall the meaning and significance of RPS lights and alarms.

NUREG 1123 KA Catalog Rev. 2 212000 RPS 212000 A3. Ability to monitor automatic operations of the REACTOR PROTECTION SYSTEM including:

212000 A3.04 System status lights and alarms 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

ILT 2020 Final Version Page: 85 of 259 Question 33 Approved View

34 K/A Importance: 3.6 Points: 1.00 R34 Difficulty: 3.00 Level of Knowledge: High Source: NEW 93069 A plant startup is in progress with reactor power at the point of adding heat.

Rx period is near infinity.

Rx MODE switch is in START & HOT STBY.

Control Rod 30-31 is selected for movement.

All APRM channels are reading between 1% and 2% power.

ALL IRM channels indicate between 3/40 and 15/40 on Range 7.

Alarm 3D64, IRM DOWNSCALE, is subsequently received.

Which of the following should the P603 Operator perform?

A. Place the Rx MODE Switch in RUN.

B. Range the alarming IRM(s) to Range 6.

C. Notch Control Rod 30-31 OUT then range IRM(s) accordingly.

D. Insert IRM Detector(s), for alarming IRM(s), until 3D64 is clear.

Answer: B ILT 2020 Final Version Page: 86 of 259 Question 34 Approved View

Answer Explanation:

23.603, IRM operating procedure precaution 3.3 states...

"During Reactor startup while power is ascending through the Intermediate range and a heatup rate is being established and/or maintained, avoid ranging down IRMs unless 3D64, IRM DOWNSCALE is received for IRM downscale. OE has shown that an IRM Upscale and half scram is possible if the IRM is ranged down too soon, or the switch is inadvertently moved two range positions."

GOP 22.000.02, Rx Startup to 25% power requires operators to adjust IRM range switches to maintain indication between 8/40-24/40 on odd ranges, and between 25/125-75/125 on even ranges.

Therefore, with the reactor near the POAH, and IRMs indicating below the required band, the operator should range DOWN IRMs to the next lower range to bring indication within the required indicating band.

Additionally, down-ranging should be performed when 3D64 is in alarm to avoid upscale trips by ranging too soon.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Placing the Mode Switch in RUN would clear the IRM downscale alarm(s) and allow the IRMs to be withdrawn (bypassing all IRM related Rod Blocks and Scrams). However, the candidate should recognize that, with the APRMs below 5%, placing the Mode Switch in RUN will result in a Rod Block due to APRM Downscale.

C. Notching out the currently selected control rod is correct to establish a heatup rate, and it would potentially clear the IRM downscale alarm(s). However, the candidate should recognize that, with the IRM downscale alarm, Control Rod withdrawal is blocked.

D. Inserting IRM(s), if withdrawn, would potentially clear IRM downscale alarm(s). However, with the Mode Switch out of Run, IRM withdrawal is prohibited (both procedurally and via interlocks). If an IRM was withdrawn, additional alarms and rod blocks would be in.

Reference Information:

23.603, INTERMEDIATE RANGE MONITORING SYSTEM.

NUREG 1123 KA Catalog Rev. 2 215003 IRM System 215003 A4. Ability to manually operate and/or monitor in the control room:

215003 A4.03 IRM range switches 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (6) Design, components, and function of reactivity control mechanisms and instrumentation.

10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO Associated objective(s):

ILT 2020 Final Version Page: 87 of 259 Question 34 Approved View

35 K/A Importance: 3.9/4.0 Points: 1.00 R35 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 90688 Which reactor power monitoring system(s) scram function is(are) designed to initiate an automatic reactor scram ONLY when the neutron monitoring shorting links are REMOVED?

A. APRM B. IRM ONLY C. SRM ONLY D. SRM AND IRM Answer: C ILT 2020 Final Version Page: 88 of 259 Question 35 Approved View

Answer Explanation:

The shorting links are removed when the plant is in MODE 5. Their removal enables the SRM instrument scram function and aligns the RPS logic for neutron monitoring such that the coincidence feature is removed, and only one channel is required to initiate a scram.

Applicants must recall the SRM reactor scram design function is the only one listed that is active ONLY when the shorting links are removed.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Incorrect since APRM scrams are active regardless of shorting link status and are only bypassed by the APRM bypass switch. Plausible because there is often confusion and misconceptions regarding the effects of installation and removal of shorting link due to the multiple actions created by their installation and removal. Also plausible because the candidate could fail to recall the normal status (INSTALLED) of the SRM shorting links.

B. Incorrect since IRM scram signals are enabled regardless of shorting link status and are only bypassed by the reactor mode switch. Plausible because there is often confusion and misconceptions regarding the effects of installation and removal of shorting link due to the multiple actions created by their installation and removal. Also plausible because the candidate could fail to recall the normal status (INSTALLED) of the SRM shorting links D. Incorrect since IRM scram signals are enabled regardless of shorting link status and are only bypassed by the reactor mode switch or IRM bypass switch. Plausible because there is often confusion and misconceptions regarding the effects of installation and removal of shorting link due to the multiple actions created by their installation and removal.

Reference Information:

SOP 23.602 This question matches the selected K/A because RO applicants must recall SRM purpose and function and the conditions under which they are active.

NUREG 1123 KA Catalog Rev. 2 215004 SRM System G2.1.27 Knowledge of system purpose and or function.

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

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO ILT 2020 Final Version Page: 89 of 259 Question 35 Approved View

Associated objective(s):

Source Range Monitoring Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Explain the purpose and location of major Source Range Monitoring System components and equipment.

ILT 2020 Final Version Page: 90 of 259 Question 35 Approved View

36 K/A Importance: 3.3/3.3 Points: 1.00 R36 Difficulty: 4.00 Level of Knowledge: Higher Source: BANK: FERMI 2010 90767 NRC EXAM APRM 2 has the following number of LPRM inputs:

A level - 5

B level - 6

C level - 4

D level - 3 Which alarm(s), if any, would result from this condition?

A. NONE B. 3D103 "APRM TROUBLE" ONLY C. 3D103 APRM TROUBLE AND 3D99 APRM INOP D. 3D103 APRM TROUBLE AND 3D113 CONTROL ROD WITHDRAWL BLOCKED Answer: D ILT 2020 Final Version Page: 91 of 259 Question 36 Approved View

Answer Explanation:

APRM 2 has less than 20 total LPRM inputs resulting in an APRM trouble alarm 3D103.

Per ARPs 3D103 and 3D113, this will result in RMCS initiating a control rod withdraw block signal to RMCS based on too few inputs and will energize annunciator 3D113.

A. is incorrect because <20 LPRM inputs will cause 3D103, APRM Trouble. This is plausible because, since each axial level has >/= 3 LPRM inputs, the candidate could see that and fail to recall the requirement to also have >/= 20 inputs total and thus determine that no alarms will occur.

B is incorrect because a rod block will also be initiated. Plausible if applicants do not recall that receipt of 3D103, APRM Trouble, will generate a rod block for any condition OTHER THAN a non-critical self-test fault.

C is incorrect because an APRM INOP signal will NOT occur, and a rod block WILL be initiated. Plausible because it is a common misconception that an INOP will accompany a TROUBLE condition; and if applicants do not remember that a rod block will be initiated by the trouble condition.

References:

SOP 23.605, APRM System ARP 3D103, APRM Trouble ARP 3D99, APRM INOP ARP 3D113, Control Rod Withdraw Block This question matches the selected K/A because RO applicants must recall physical connections and cause effect relationship between an APRM channel and RMCS, under APRM trouble conditions that cause a rod withdraw block signal to RMCS.

NUREG 1123 KA Catalog Rev. 2 215005 APRM/LPRM 215005 K1. Knowledge of the physical connections and/or cause effect relationships between AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM and the following:

215005 K1.10 3.3/3.3 Reactor manual control system: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILO 2010 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

Power Range Monitoring and Rod Block Monitoring Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: List the interlocks associated with Power Range Neutron Monitoring System components.

ILT 2020 Final Version Page: 92 of 259 Question 36 Approved View

37 K/A Importance: 2.7/2.8 Points: 1.00 R37 Difficulty: 3.00 Level of Knowledge: Source: MODIFIED: FERMI 90248 Fundamental 2019 RETAKE R39 (#87853)

A loss of RCIC Logic Bus A will result in a loss of...

A. automatic RCIC initiation capability.

B. automatic RCIC turbine trip capability.

C. isolation capability for E5150-F007, RCIC Stm Line Inbd Iso Valve ONLY.

D. isolation capability for E5150-F007, RCIC Stm Line Inbd Iso Valve AND E5150-F008, RCIC Stm Line Otbd Iso Valve.

Answer: A & B ILT 2020 Final Version Page: 93 of 259 Question 37 Approved View

Answer Explanation: A Note: This question has been significantly modified from the 2019 NRC Retake version, Q39 on that exam, by modifying conditions in the stem of the question therefore resulting in a different correct response. This version of the question asks for the impact of a loss of Logic Bus A while the previous version asked for the impact of a loss of power to the RCIC Inverter ONLY. The RCIC Inverter is powered by 2PA2-5, Pos 9. Loss of power to the RCIC Inverter is limited and only impacts indications for pump suction/discharge pressure and turbine supply/exhaust pressure. Also, the inverter supplies power to E51-K615, RCIC Moore Controller, which would disable the ability to automatically control RCIC flow (previous correct answer). This version asks for the impact of loss of Logic Bus A, which is much more substantial and would prevent automatic system initiation capability (current correct answer).

Per ARP 1D56, RCIC LOGIC BUS POWER FAILURE:

Loss of A Bus:

- RCIC will not auto start.

- E5150-F008, RCIC Stm Line Otbd Iso Vlv, will not auto isolate.

- E5150-F062, RCIC Exh Vac Bkr Otbd Iso Vlv, will not auto isolate.

- RCIC will not isolate with RCIC Logic A Manual Isolation Pushbutton.

- RCIC Suction will not shift to Torus on Low CST Level.

- E5150-F010, RCIC Pump CST Suction Iso Valve, will not auto open or auto close.

- E5150-F045, RCIC Turb Steam Inlet Vlv, closes, if open.

- E5150-F013, RCIC Disch To Fw Inbd Iso Valve, closes, if open.

- E5150-F095, RCIC Turb Stm Inlet Byp Vlv, closes, if open.

Loss of B Bus:

- E5150-F007, RCIC Stm Line Inbd Iso Vlv, will not auto isolate.

- E5150-F084, RCIC Exh Vac Bkr Inbd Iso Vlv, will not auto isolate.

- RCIC L-8 and Isolation Logic B trip will not function.

B is also correct since the loss of logic bus A will prevent energizing the K8 relay which closes a contact in the RCIC remote trip circuit [which is powered from logic bus B] thereby removing the automatic trip capability for RCIC.

Reference:

1D56, RCIC Logic Bus Power Failure.

I-2235-01, RCIC power supply schematic Distractor explanation:

C is incorrect since this results from a loss of logic power B, and will be unaffected by loss of logic bus A.

Plausible if applicants incorrectly recall RCIC power supplies.

D is incorrect since this results from a loss of logic power B, and will be unaffected by loss of logic bus A.

Plausible if applicants incorrectly recall RCIC power supplies.

K/A Match Justification:

This question matches the selected k/a since RO applicants must recall the power supply to the RCIC auto initiation logic without reference to procedures.

ILT 2020 Final Version Page: 94 of 259 Question 37 Approved View

NUREG 1123 KA Catalog Rev. 2 217000 RCIC System 217000 K2 Knowledge of electrical power supplies to the following:

217000 K2.02 2.8*/2.9* RCIC initiation signals (logic) 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental Modified RO Associated objective(s):

Reactor Core Isolation Cooling Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Describe the normal and alternate power supplies to RCIC System components.

ILT 2020 Final Version Page: 95 of 259 Question 37 Approved View

38 K/A Importance: 4.4/4.4 Points: 1.00 R38 Difficulty: 4.00 Level of Knowledge: Higher Source: BANK: FERMI 2006 91330 RETAKE R48 A plant transient has occurred. All MSIVs are closed, and HPCI is operating and injecting into the vessel. 1D57, ADS/SRV/EECW TCV POWER SUPPLY FAILURE, has alarmed, and investigation reveals that 2PA2-5 Circuit 1 is de-energized and cannot be restored.

The following plant conditions are noted:

RPV Water Level .............................. 35 inches (slowly lowering)

Reactor pressure .............................. 900 psig (slowly lowering)

Drywell pressure ............................... 1.0 psig (slowly rising)

About 3 minutes later, the following H11-P601 panel annunciators alarm:

1D31, ADS DRYWELL PRESS BYPASS TIMER INITIATE A / B LOGIC

1D36, ADS ECCS PUMP CH B PERMISSIVE From the receipt of alarms 1D31 and 1D36, the EARLIEST that ADS Valves should be expected to AUTOMATICALLY OPEN is _____________.

A. 105 seconds B. 7 minutes C. 8 minutes 45 seconds D. Never (valves remain closed)

Answer: C ILT 2020 Final Version Page: 96 of 259 Question 38 Approved View

Answer Explanation:

C. Annunciators indicate that L1 (31.8 inches) has occurred.

L1 + 7 minutes (DW Bypass Timer) = 7 minutes + 105 seconds = 8 minutes and 45 seconds ADS Valves OPEN. Loss of 2PA2-5 disables ADS Logic A, but ADS Logic B is powered and has dual power supplies (2PA2-5 & 2PA2-6). ADS Logic B is sufficient to initiate ADS.

A is plausible, 105 seconds from L1, IF Hi DW Pressure existed, this would be correct.

B is plausible, 7 minutes from L1, candidate may select if misconception exists about logic and 105 sec timer.

D is plausible, if candidate believes loss of 2PA2-5 will disable both logic trains/systems.

References:

1D57, ADS/SRV/EECW TCV POWER SUPPLY FAILURE 1D31, ADS DRYWELL PRESS BYPASS TIMER INITIATE A/B LOGIC 1D36, ADS ECCS PUMP CH B PREMISSIVE NUREG 1123 KA Catalog Rev. 2 218000 ADS 218000 K3. Knowledge of the effect that a loss or malfunction of the AUTOMATIC DEPRESSURIZATION SYSTEM will have on the following:

218000 K3.01 4.4*/4.4* Restoration of reactor water level after a break that does not depressurize the reactor when required 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILO 2006 Retake Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

Automatic Depressurization System Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Describe general ADS operation, including component operating sequence, normal operating parameters, and expected system response.

ILT 2020 Final Version Page: 97 of 259 Question 38 Approved View

39 K/A Importance: 3.3/3.7 Points: 1.00 R39 Difficulty: 2.00 Level of Knowledge: Fund Source: NEW 90847 The plant has experienced an ATWS. Reactor Power is 100%. The crew has entered 29.100.01, RPV Control.

The CRS has directed operators to perform ATWS actions.

Which of the following isolation signals (if any) must be bypassed during these actions?

A. NO isolation signals B. ALL MSIV isolation signals ONLY C. RPV Level 1 MSIV isolation signal ONLY D. ALL MSIV AND Main Steam Line Drain Valve isolation signals Answer: C ILT 2020 Final Version Page: 98 of 259 Question 39 Approved View

Answer Explanation:

Per 29.100.01, RPV Control, step RC-3 directs ATWS actions be performed. This set of steps directs the use of 29.ESP.11, Defeat of RPV Level 1 MSIV Isolation Signals, in which only MSIV L1 isolation signals are bypassed to facilitate lowering RPV water level.

A is incorrect since the L1 signal must be bypassed. Plausible if applicants dont recall ATWS actions.

B is incorrect since only the L1 signal is bypassed. Plausible since all MSIV isolation signals are bypassed under some EOP conditions.

D is incorrect is incorrect since only the L1 signal is bypassed. Plausible since all MSIV & steam line drain valve isolation signals are bypassed under some EOP conditions.

Reference Information:

29.ESP.12, Defeat of All MSIV And Main Steam Line Drain Valve Isolation Signals 29.ESP.11, Defeat of RPV Level 1 MSIV Isolation Signals This question matches the selected K/A because RO applicants must recall details of the MSIV isolation manual defeat actions used during emergency procedures.

NUREG 1123 KA Catalog Rev. 2 223002 PCIS/NSSS 223002 K4. Knowledge of PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF design feature(s) and/or interlocks which provide for the following:

223002 K4.08 Manual defeating of selected isolations during specified emergency conditions 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

Emergency Support Procedures Performance Enabler Defeat all Main Steam Isolation Valve and Main Steam Line Drain Valve Isolation Signals ILT 2020 Final Version Page: 99 of 259 Question 39 Approved View

40 K/A Importance: 3.3/3.5 Points: 1.00 R40 Difficulty: 2.00 Level of Knowledge: Higher Source: BANK: CGS 2011 90769 Cognitive Level NRC The crew is flooding the RPV using RHR with suction from the Torus per 29.100.01, sheet 3, RPV Flooding.

Which ONE of the following is a valid indication the crew can use to confirm that the RPV has been flooded up to the Main Steam Lines?

A. Torus water level rising.

B. Torus water temperature rising.

C. Tailpipe temperatures of open SRVs rising.

D. Tailpipe temperatures of open SRVs lowering.

Answer: D ILT 2020 Final Version Page: 100 of 259 Question 40 Approved View

Answer Explanation:

NOTE: BANK source for this question is the 2011 CGS NRC Exam.

Per ODE-10:

Some combination of the following indications can be used as indication that the RPV has been flooded to the Main Steam Lines:

Increasing RPV pressure (after initially lowering due to depressurization) as noncondensables are compressed.

Tailpipe temperatures of open SRVs decreasing to subcooled values.

If injection sources are aligned with suction from the Torus, Torus water level decreases as the RPV and Steam Lines are flooded, then stabilizes when the steam lines are full.

Actuation of Main Steam Line, HPCI or RCIC high steam flow logic, if they are not

isolated.

Water leakage from HPCI or RCIC turbine shaft seals, if they are not isolated.

HPCI or RCIC steam line drain pot high level alarm, if they are not isolated.

A is incorrect because torus W/L will lower as the vessel is flooded, then stabilize once the MSLs are reached. Plausible if applicants believe that some leakage back to the torus would occur once flooded conditions are reached. Also plausible because this indication would be valid if injection was occurring from an external source (outside containment).

B is incorrect because torus W/T changes are not used to verify flooding has occurred. Plausible since torus temp may rise due to SRV lift while depressurizing, however, the value or trend of torus temp cannot provide confirmation of flooded conditions.

C is incorrect because tailpipe temperature will lower. Plausible since rising tailpipe temps are typically used to confirm SRVs are open.

References:

29.100.01 sheet 3, RPV Flooding ODE-10, EOP Expectations This question matches the selected K/A because RO applicants recall the operational implications of SRV tailpipe temperature changes under flooding conditions NUREG 1123 KA Catalog Rev. 2 239002 SRVs 239002 K5. Knowledge of the operational implications of the following concepts as they apply to RELIEF/SAFETY VALVES :

239002 K5.04 Tail pipe temperature monitoring 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO ILT 2020 Final Version Page: 101 of 259 Question 40 Approved View

Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Performance Enabler Execute steps of Reactor Pressure Vessel Flooding (RF)

ILT 2020 Final Version Page: 102 of 259 Question 40 Approved View

41 K/A Importance: 3.1/3.1 Points: 1.00 R41 Difficulty: 3.00 Level of Knowledge: Low Source: BANK 90307 The plant is operating at 100% power with the following:

RPV Level is controlling at 197.

Feedwater Control System (FWCS) is in 3 element control.

The 'A' RPV Level signal is selected as the lead.

Which of the following will force the Feedwater Control System to Single Element Control?

Failure of the ...

A. 'A' Level input ONLY.

B. 'A' Feedwater Flow input.

C. 'A' Steam Line Flow input.

D. BOTH the 'A' and 'B' Level inputs.

Answer: B ILT 2020 Final Version Page: 103 of 259 Question 41 Approved View

Answer Explanation:

Per 23.107, Section 1.1 System Description, Failure of a single feedwater flow signal or two main steam line flow signals will cause DCS logic to automatically force 1 element RPV water level control.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could confuse the terminology in 23.107 and determine that Level control will shift to Single Element if the lead level input fails. This is incorrect because, if the lead selected RPV water level signal fails, DCS logic will automatically transfer RPV water level control to the median signal from the three remaining RPV water level signals, preventing an RPV water level transient. Median water level control is not the same as 1 element water level control.

C. The candidate could determine that failure of 1 steam flow input will cause feedwater control to transfer to single element, which is plausible because feedwater DCS does respond to failure of one steam flow input, just not in this way. Failure of a single main steam line flow signal will cause DCS logic to substitute the median flow signal from the other 3 steam lines, which is not the same as forcing 1 element water level control.

D. The candidate could determine that failure of both the A and B level inputs (the ONLY level inputs, of the 4, that can be selected as the Lead RPV Water Level signal), is enough to force 1 element water level control. However, if a second level signal fails, DCS will operate correctly using the median level signal of the two remaining signals, IF one or both of the level signal failures was a lead selected water level signal. If both of level signal failures were NOT lead selected signals, DCS will continue to operate on the selected lead signal. Therefore, this distractor is incorrect since it will not force 1 element control.

Reference Information:

23.107, Reactor Feedwater and Condensate Systems.

Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 259002 Reactor Water Level Control System 259002 K6 Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR WATER LEVEL CONTROL SYSTEM:

259002 K6.04 3.1/3.1 Reactor feedwater flow input 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental RO ILT 2020 Final Version Page: 104 of 259 Question 41 Approved View

Associated objective(s):

Feedwater Control Cognitive Terminal In accordance with approved plant procedures/references, under all conditions of the Feedwater Control System: Describe system operation, including component operating sequence, normal operating parameters, and expected system response.

ILT 2020 Final Version Page: 105 of 259 Question 41 Approved View

42 K/A Importance: 2.9/3.1 Points: 1.00 R42 Difficulty: 3.00 Level of Knowledge: Low Source: BANK SOURCE: 2006 90327 FERMI ILT EXAM During operation of Div 2 SGTS, the following occurs:

17D9, DIV II SGTS AIR FLOW STOPPED, alarms.

T46-R800B, Div 2 SGTS Exh Gas Flow Recorder, indicates 1800 scfm.

These conditions will result in which of the following automatic actions?

A. Trip of the Div 2 SGTS Heater.

B. Trip of the Div 2 SGTS Exhaust Fan.

C. Automatic start of the Div 1 SGTS Exhaust Fan.

D. Automatic start of the Div 2 SGTS Train Cooling Fan.

Answer: A ILT 2020 Final Version Page: 106 of 259 Question 42 Approved View

Answer Explanation:

Per 17D9, the candidate must recall that the low flow condition presented in the stem of the question will result in trip of the Div 2 SGTS heater to prevent over-heating of the heater element due to low flow.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The candidate could conclude that low flow trips the Exhaust fan, which is how other pumps/fans respond for other systems, such as the RWCU system pumps that trip on low flow. This distractor is incorrect because the SGTS Exhaust fan does not trip on a low flow condition.

C. The candidate could conclude that the non-running division of SGTS will auto start because of either (1) directly due to the low flow condition of the running SGTS division or (2) indirectly because of degrading Reactor Building Differential Pressure due to the low flow condition. This distractor is incorrect because SGTS does not auto start on low flow or high RB differential pressure.

D. The candidate could conclude that the normally non-running cooling fan will auto on low SGTS flow to prevent over-heating of the SGTS charcoal bed due to the low flow condition. This is incorrect because the Cooling fan auto starts on high charcoal bed temperature and not low system flow.

Reference Information:

17D9, DIV II SGTS AIR FLOW STOPPED Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 261000 Standby Gas Treatment System 261000 A1 Ability to predict and/or monitor changes in parameters associated with operating the STANDBY GASTREATMENT SYSTEM controls including:

261000 A1.01 2.9/3.1 System flow 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Fermi 2 NRC Exam Usage ILO 2006 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental RO Associated objective(s):

Standby Gas Treatment System Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: List the interlocks associated with Standby Gas Treatment System components.

ILT 2020 Final Version Page: 107 of 259 Question 42 Approved View

43 K/A Importance: 2.7/2.9 Points: 1.00 R43 Difficulty: 4.00 Level of Knowledge: High Source: NEW 90827 The plant is operating at 100% power with all electrical busses and distribution cabinets powered from their normal power supplies.

Multiple alarms occur on multiple control room panels.

Upon evaluating the H11-P810 Panel, the CRLNO observes that Bus 65D has been de-energized.

(1) Among others, which 480V bus has been impacted by bus 65D being de-energized?

(2) Which section of AOP 20.300.MPU will be used to mitigate the consequences of bus 65D being de-energized?

A. (1) 72A.

(2) MPU #5.

B. (1) 72A.

(2) MPU #6.

C. (1) 72L.

(2) MPU #5.

D. (1) 72L.

(2) MPU #6.

Answer: D ILT 2020 Final Version Page: 108 of 259 Question 43 Approved View

Answer Explanation:

The candidate must recall that de-energizing bus 65D will, in turn, de-energize bus 65L, which in turn will de-energize MPU #6 (normally energized via MCC 72L) due to Bus 72L being energized.

This should lead the candidate to determine that (1) Bus 72L has been impacted and (2) the MPU#6 section of 20.300.MPU must be entered and performed.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. (1) 72A is plausible because it is a 480V BOP bus that is of similar importance to Bus 72L (for example, both are the normal AND alternate power sources for BOP MPUs 5 and 6). Also, since the stem of the question indicates a loss of Bus 6D, which in turn de-energizes 65D, there is no direct cue that Bus 72L would be impacted. The operator could therefore incorrectly recall 4160V / 480V electrical distribution and determine that Bus 72A is impacted. However, loss of Bus 65D (65L) impacts 480V Bus 72L and NOT Bus 72A. (2) MPU #5 is plausible because 480V Bus 72A is the normal power supply to MPU #5.

B. (1) 72A is plausible because it is a 480V BOP bus that is of similar importance to Bus 72L (for example, both are the normal AND alternate power sources for BOP MPUs 5 and 6). Also, since the stem of the question indicates a loss of Bus 6D, which in turn de-energizes 65D, there is no direct cue that Bus 72L would be impacted. The operator could therefore incorrectly recall 4160V / 480V electrical distribution and determine that Bus 72A is impacted. However, loss of Bus 65D (65L) impacts 480V Bus 72L and NOT Bus 72A. (2) MPU #6 is plausible because the AOPs for loss of Bus 65D and 65L both direct the operators to 20.300.MPU for MPU #6.

C. (1) Part 1 is correct. (2) MPU #5 is plausible because the candidate could confuse the normal power supplies to MPUs 5 and 6. Also, 480V Bus 72R would also be de-energized upon loss of Bus 65D (65L) and Bus 72R is the alternate power supply for MPU #5.

Reference Information:

SD-2500-01, One Line Diagram Plant 4160 and 480V System Service.

20.300.65D, Loss of 65D AOP.

20.300.MPU, Loss of MPU.

20.300.65L, Loss of 65L AOP.

Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 262001 AC Electrical Distribution 262001 A2. Ability to predict (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations:

262001 A2.06 2.7/2.9 Deenergizing a plant bus 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO ILT 2020 Final Version Page: 109 of 259 Question 43 Approved View

Associated objective(s):

4160/480V Electrical Distribution Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: Identify abnormal and emergency operating procedures associated with the 4160/480V Electrical Distribution System.

ILT 2020 Final Version Page: 110 of 259 Question 43 Approved View

44 K/A Importance: 2.8/3.1 Points: 1.00 R44 Difficulty: 3.00 Level of Knowledge: High Source: NEW 90707 How will the UPS system respond to a loss of 480VAC bus 72R?

A. UPS A loads will be automatically supplied by the UPS B rectifier.

B. UPS B loads will be automatically supplied by the UPS A rectifier.

C. UPS A loads will be automatically transferred to its alternate supply regulator.

D. UPS B loads will be automatically transferred to its alternate supply regulator.

Answer: B ILT 2020 Final Version Page: 111 of 259 Question 44 Approved View

Answer Explanation:

Per the UPS system drawing above (located in 23.308.01):

The examinee must recall the flow of power through the UPS system. The examinee must recall that the normal supply to UPS B is 72R, and the alternate supply is either the A rectifier, the UPS Battery, or bus 72M depending upon the mode of failure. In this instance, the examinee must recall that the normal supply to UPS B is lost resulting in the UPS B inverter being automatically supplied by the UPS A rectifier.

ILT 2020 Final Version Page: 112 of 259 Question 44 Approved View

UPS B will only transfer to 72M if the UPS B inverter fails. It will only be supplied by the UPS battery if the normal supply to BOTH UPS A and B are lost. Neither of these events have occurred in this case.

A is incorrect UPS A loads will remain supplied by the UPS A inverter and rectifier following loss of 72R.

This is plausible if applicants confuse the normal AC supplies to UPS A and B.

C incorrect because UPS A normal supply is bus 72M, and UPS A loads will remain supplied by the UPS A inverter and rectifier following loss of 72R. The only event that would transfer A loads to the alternate supply regulator output is failure of the A inverter. This is plausible if applicants confuse the normal AC supplies to UPS A and B.

D is incorrect because UPS B loads will only be supplied by the regulator following a failure of the UPS B inverter. This is plausible since this automatic action will occur under some conditions.

References:

SOP 23.308.01, Uninterruptible Power Supply System AOP 20.300.72R, Loss of Bus 72R This question matches the selected K/A because RO applicants must recall the response of the UPS system to a loss of normal AC supply and the resultant automatic transfer its alternate supply.

NUREG 1123 KA Catalog Rev. 2 262002 UPS (AC/DC) 262002 A3. Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including:

262002 A3.01 2.8/3.1 Transfer from preferred to alternate source 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO Associated objective(s):

UPS Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: Describe Uninterruptible Power Supply System automatic features.

ILT 2020 Final Version Page: 113 of 259 Question 44 Approved View

45 K/A Importance: 3.3 Points: 1.00 R45 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 92389 Loss of which battery would result in loss of indicating lights for breakers on 4160V Busses 64B and 11EA and 480V Busses 72B and 72EA?

A. 2A-1 B. 2A-2 C. 2B-1 D. 2B-2 Answer: A ILT 2020 Final Version Page: 114 of 259 Question 45 Approved View

Answer Explanation:

Div 1 Battery 2A-1 supplies Circuit Breaker Control Power to 4160V Busses 64B and 11EA and 480V Busses 72B and 72EA. Loss of this battery would cause loss of indicating lights for breakers on these busses.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. This distractor would be true for 4160V busses 64C and 12EB and 480V Busses 72C and 72EB.

However, the busses listed in the stem receive control power from battery 2A-1.

C. This distractor would be true for 4160V busses 65E and 13EC and 480V Busses 72E and 72EC.

However, the busses listed in the stem receive control power from battery 2A-1.

D. This distractor would be true for 4160V busses 65F and 14ED and 480V Busses 72F and 72ED.

However, the busses listed in the stem receive control power from battery 2A-1.

Reference Information:

20.300.260VESF NUREG 1123 KA Catalog Rev. 2 263000 DC Electrical Distribution 263000 A4. Ability to manually operate and/or monitor in the control room:

263000 A4.01 Major breakers and control power fuses: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

DC Electrical Distribution Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: Describe the impact on plant operations of a loss of the following DC buses, and describe the actions necessary to correct, control or mitigate the loss of DC power: 260/130VDC ESF bus, 260/130VDC BOP bus, 48/24VDC Bus ILT 2020 Final Version Page: 115 of 259 Question 45 Approved View

46 K/A Importance: 3.1/4.2 Points: 1.00 R46 Difficulty: 3.00 Level of Knowledge: Higher Source: BANK 90529 With the plant operating at 80% power, at 0800 on August 28, EDG11 is declared INOPERABLE to perform emergent maintenance.

Which ONE of the following describes the LATEST TIME that SR 3.8.1.1 "Verify correct breaker alignment and indicated power availability for each offsite circuit" can be completed WITHOUT entering a condition which requires a unit shutdown?

A. 0815 on August 28 B. 0850 on August 28 C. 0905 on August 28 D. 0915 on August 29 Answer: B ILT 2020 Final Version Page: 116 of 259 Question 46 Approved View

Answer Explanation:

SR 3.8.1.1 is due WITHIN ONE HOUR, 0850 August 28 is the latest time which complies.

Distracter Explanation:

A. Is incorrect and plausible, but NOT the latest time (15 minutes for event classification)

B. Is incorrect and plausible, SR 3.8.1.1 expired at 0900 - NO 1.25 extensions are permitted on initial period D. Is incorrect and plausible, SR 3.8.1.1 expired at 0900 - NO 1.25 extensions are permitted on initial period Reference Information:

T.S. 3.8.1 (pg 3.8-1 to 3.8-9)

NUREG 1123 KA Catalog Rev. 2 G2.2.36 Ability to analyze the effect of maintenance activities such as degraded power sources, on the status of limiting conditions for operations Technical Specifications 3.8.1 AC Sources Operating 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILO 2015 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

Emergency Diesel Generator Cognitive Terminal In accordance with approved plant procedures/references, given the operating conditions and parameters for the Emergency Diesel Generator System: Identify the system related Technical Specification operability requirements and bases, with an emphasis on action statements requiring actions within an hour ILT 2020 Final Version Page: 117 of 259 Question 46 Approved View

47 K/A Importance: 3.1/3.2 Points: 1.00 R47 Difficulty: 3.00 Level of Knowledge: High Source: BANK 90508 The plant is operating at 100% power when the following occur:

7D54, INTERRUPTIBLE CONTROL AIR HEADER PRESS LOW, alarms.

P50-R870, IAS Header Pressure Indicator, reads 80 psig (lowering).

If this trend continues, which of the following describes the component(s) affected, AND the correct failure mode(s), which will require the Reactor MODE Switch to be placed in SHUTDOWN?

A. An Inboard MSIV begins to CLOSE.

B. An Outboard MSIV begins to CLOSE.

C. A Control Rod drifts in due to C11-F002A, CRD Flow Control Valve, failing OPEN.

D. A Control Rod Drive Hydraulic Pump TRIPS on Low Suction Pressure due to C11-F412, CRD Pump Suction Pressure Control Valve, failing SHUT.

Answer: B ILT 2020 Final Version Page: 118 of 259 Question 47 Approved View

Answer Explanation:

Per 20.129.01, Bases for Override 1 in 20.129.01, IAS supplies air to the Outboard MSIVs. If IAS is lost to the outboard MSIVs, they would eventually drift closed as the pressure is bled off in the supply lines, causing a reactor scram. Therefore, the candidate must first recall that IAS supplies air to the Outboard MSIV and second the candidate must recall that loss of air pressure will cause them to close.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The Inboard MSIVs are supplied air to operate and will close on a loss of air pressure. However, the Inboard MSIV are not supplied by IAS, rather they are supplied air to hold them open by Division 1 NIAS, making this distractor incorrect. .

C. Control rods drifting inward would require the Mode Switch be taken to Shutdown and control rods would drift in, on a loss of IAS, due to loss of air to the scram inlet and outlet valves. Also, opening the CRD FCV could raise under-piston pressure and result in Control Rods drifting inwards, which is accomplished during alternate control rod insertion during an ATWS. However, the CRD Flow Control Valve fails CLOSED on a loss of IAS which would not cause control rods to drift.

D. If both CRD pumps were tripped, due to low suction pressure caused by loss of IAS, the Mode Switch would eventually have to be placed in shutdown because of the impact on Control Rod Scram Accumulator pressure. However, when the C11-F412 shuts on loss of air pressure, CRD pump suction will automatically transfer to the CST via the C1100-F212, CST to CRD Pumps Secondary Supply Check Valve, maintaining CRD pump suction and preventing a pump trip.

Reference Information:

20.129.01, Loss of Station and/or Control Air AOP.

20.129.01 - BASES.

Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 300000 Instrument Air System 300000 K1. Knowledge of the connections and/or cause-effect relationships between INSTRUMENT AIR SYSTEM and the following:

300000 K1.05 3.1/3.2 Main Steam Isolation Valve air 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (3) Mechanical components and design features of reactor primary system.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

Compressed Air Systems Cognitive Terminal In accordance with approved plant procedures/references, given various controls and indications for operation of the Compressed Air System: Discuss the system interrelationships with other plant systems.

ILT 2020 Final Version Page: 119 of 259 Question 47 Approved View

48 K/A Importance: 2.9 Points: 1.00 R48 Difficulty: 2.00 Level of Knowledge: Fund Source: BANK 91890 With the plant operating at 100% power the following occurs:

9D10, DIV 1 480 V ESS BUS 72C BKR TRIPPED, alarms.

CMC Switch for BUS 64C POS C11, 4160V FEED TO BUS 72C, indicates TRIPPED.

Which of the following describes the impact on the Reactor Building Closed Cooling Water (RBCCW) Pumps?

A. ONLY P4200-C001, North RBCCW Pump, has lost power.

B. ONLY P4200-C003, South RBCCW Pump, has lost power.

C. BOTH P4200-C001 AND P4200-C002, North and Center RBCCW Pumps, have lost power.

D. BOTH P4200-C003 AND P4200-C002, South and Center RBCCW Pumps, have lost power.

Answer: A ILT 2020 Final Version Page: 120 of 259 Question 48 Approved View

Answer Explanation:

Per 23.127, RBCCW/EECW SOP, Attachment 2C, RBCCW Electrical Lineup, Bus 72C powers ONLY P4200-C001, North RBCCW Pump.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. is plausible; would be true for a loss of Bus 72E.

C. is plausible; would be true for a loss of Buses 72C and 72 F. Having 2 pumps powered from the same 480V bus is also plausible because two TBCCW Pumps (Center and South) are powered from 480V Bus 72N.

D. is plausible; would be true for a loss of Buses 72E and 72 F. Having 2 pumps powered from the same 480V bus is also plausible because two TBCCW Pumps (Center and South) are powered from 480V Bus 72N.

Reference Information:

23.127, RBCCW/EECW SOP.

Plant Procedures 23.127 NUREG 1123 KA Catalog Rev. 2 400000 Component Cooling Water System 400000 K2. Knowledge of electrical power supplies to the following:

400000 K2.01 CCW pumps 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental RO Associated objective(s):

Reactor Building Closed Cooling Water/Emergency Equipment Cooling Water Cognitive Terminal In accordance with approved plant procedures/references, under all conditions of the RBCCW/EECW System: Identify the normal and alternate power supplies to system components.

ILT 2020 Final Version Page: 121 of 259 Question 48 Approved View

49 K/A Importance: 3.7 Points: 1.00 R49 Difficulty: 3.00 Level of Knowledge: High Source: NEW 92609 The plant has recently entered MODE 4.

The following conditions exist:

RPV Water Level is 230".

RPV Temperature is 120°F.

A loss of Shutdown Cooling has occurred.

Shutdown Cooling cannot be restored to either loop of RHR.

There is no requirement to maintain Reactor Coolant Temperature 200°F and nothing would prevent a MODE change.

Which of the following would restore reliable reactor coolant temperature monitoring?

A. Raise RPV water level to 255".

B. Place a Reactor Recirculation Pump in service.

C. Maximize Recirculation Loop Flow through RWCU.

D. Line up to bleed steam to the Main Condenser or Torus.

Answer: B ILT 2020 Final Version Page: 122 of 259 Question 49 Approved View

Answer Explanation:

Per 22.000.05 Attachment 3, Reactor Cold Shutdown Temperature and Pressure Monitoring Data Sheet, coolant temperature is monitored using the RHR HX temperature (if RHR is in service) or Recirc Loop Temperature (if a loop is in service).

20.205.01, Loss of Shutdown Cooling, bases for Action L.1 states that, with no RHR shutdown cooling subsystem and no recirculation pump in operation, reactor coolant circulation must be restored without delay. Starting a Reactor Recirc Pump will provide the necessary circulation for monitoring coolant temperature.

Therefore, the examinee must determine that, with no SDC available, the MINIMUM action necessary to restore reliable coolant temperature monitoring is to start a RR Pump, which will allow temperature monitoring by utilizing Recirc Loop Temperatures IAW 22.000.05 Attachment 3.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. 220 to 255" is the normal shutdown level band with RHR in SDC to promote natural circulation. The examinee could conclude that raising RPV water level high in the band will promote greater natural circulation (which it will) and therefore restore reliable temperature monitoring. However, 22.000.05 and 20.205.01 are both clear that reliable temperature monitoring is obtained by establishing flow through either an RHR SDC loop or by starting a RR Pump.

C. Maximizing Bottom Head Drain Flow is an action performed, IAW 20.138.01 RR Pump Trip AOP, when both RR Pumps are tripped. This action is performed to ensure that the bottom head temperature indication is accurate for the impending restart of a reactor recirculation pump, and also limits the stratification of water in the bottom head area of the vessel. This could lead the examinee to conclude that maximizing Recirculation Loop Flow will restore accurate coolant temperature monitoring via the RR Loops. However, 22.000.05 and 20.205.01 are both clear that reliable temperature monitoring is obtained by establishing flow through either an RHR SDC loop or by starting a RR Pump and no guidance exists to maximize RR loop flow.

D. 20.205.01, Loss of Shutdown Cooling Condition L requires placing a Reactor Recirculation Pump in service, and bleeding steam to the Main Condenser or Torus, to establish coolant circulation and decay heat removal. The examinee could determine that bleeding steam would restore coolant temperature monitoring. However, 22.000.05 and 20.205.01 are both clear that reliable temperature monitoring is obtained by establishing flow through either an RHR SDC loop or by starting a RR Pump. Bleeding steam is necessary for Decay Heat Removal.

Reference Information:

20.205, Loss of Shutdown Cooling AOP and BASES.

22.000.05 Attachment 3.

20.138.01 RR Pump Trip AOP.

23.205 RHR System SOP.

ILT 2020 Final Version Page: 123 of 259 Question 49 Approved View

NUREG 1123 KA Catalog Rev. 2 205000 RHR Shutdown Cooling Mode 205000 K3. Knowledge of the effect that a loss or malfunction of the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) will have on the following:

205000 K3.04 3.7/3.7 Recirculation loop temperatures 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New NRC Early Review RO Associated objective(s):

Residual Heat Removal Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: Identify operating procedures associated with the Residual Heat Removal System.

ILT 2020 Final Version Page: 124 of 259 Question 49 Approved View

50 K/A Importance: 2.9/3.0 Points: 1.00 R50 Difficulty: 3.00 Level of Knowledge: Higher Source: NEW 90687 A plant startup is in progress with:

All SRMs reading between 5x104 and 9x104 cps.

All IRMs mid-scale on range 3.

The P603 operator selects all SRM detectors for withdrawal, but inadvertently selects IRM A also.

He then depresses and holds the Drive Out pushbutton.

IRM A detector will...

A. withdraw, then automatically stop when the Retract Permit light goes OFF.

B. NOT withdraw because the Retract Permit light will be OFF in this condition.

C. withdraw until fully retracted, and a control rod withdraw block will be initiated.

D. withdraw, then automatically stop when the detector's "not full in" limit switch is actuated.

Answer: C ILT 2020 Final Version Page: 125 of 259 Question 50 Approved View

Answer Explanation:

When the withdraw pushbutton is depressed, all detectors that are selected for movement will withdraw, and will continue to withdraw until the pushbutton is released. Once the IRM detector not full in limit switch is activated a control rod withdraw block will be initiated. This automatic feature is active when the reactor mode switch is NOT in RUN as it will be in this instance.

Therefore, the examinee must recall that there are no interlocks preventing the IRM detector from being withdrawn in any configuration, but that a rod block will be initiated once the detectors not full in limit switch is activated.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Incorrect since there are no automatic restrictions on IRM detector movement. Plausible because there is a common misconception that the retract permit light circuit will automatically block detector motion.

B. Incorrect since there are no automatic restrictions on IRM detector movement. Plausible because there is a common misconception that the retract permit light circuit will automatically block detector motion.

D. Incorrect since there are no automatic restrictions on IRM detector movement. Plausible if applicants incorrectly believe that the not full in limit switch will interrupt the detector drive circuit to stop motion.

Reference Information:

SOP 23.603 ARP 3D113 This question matches the selected K/A because RO applicants must recall IRM detector design features and interlocks related to detector movement and apply them under the given conditions.

NUREG 1123 KA Catalog Rev. 2 215003 IRM System 215003 K4. Knowledge of INTERMEDIATE RANGE MONITOR (IRM) SYSTEM design feature(s) and/or interlocks which provide for the following:

215003 K4.05 2.9/3 Changing detector position 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO ILT 2020 Final Version Page: 126 of 259 Question 50 Approved View

Associated objective(s):

Intermediate Range Monitoring Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Explain the basic principles of operation for the Intermediate Range Monitoring System and the major components and equipment.

ILT 2020 Final Version Page: 127 of 259 Question 50 Approved View

51 K/A Importance: 3.3/3.5 Points: 1.00 R51 Difficulty: 4.00 Level of Knowledge: Higher Source: BANK: FERMI 2015 90848 NRC EXAM, 2017 AUDIT The plant is operating at 95% power.

A leak develops on the H21-P004 rack causing the excess flow check valve on the VARIABLE leg of the WIDE Range instruments to close in response to the leak.

What would be the DIRECT effect on the plant from this change in sensed level?

A. RWCU isolates.

B. TIPs retract and isolate.

C. Reactor low water Level 3 scram.

D. Level 8 trip of both RFPs and Main Turbine.

Answer: A ILT 2020 Final Version Page: 128 of 259 Question 51 Approved View

Answer Explanation:

Wide range instruments on a rack means that the leak affects 2 instruments, and because the check valve is closed, the variable leg of the two instruments will be low, so that the instruments will indicate failed low. Because the rack number is even P004, and the logic is NSSS A= Channels A, B and B=Channels C,D. Therefore NSSS group isolations for the wide range instruments would occur on NSSS logic A. In this case Group 10 for RWCU is one isolation that would occur Distracter Explanation:

B is incorrect since this action comes from narrow range instruments. Plausible if student thinks group 15 is off of wide range or connects this leak to the narrow level instruments.

C is incorrect since this action comes from narrow range instruments. Plausible if student thinks there is an RPS Trip associated with the wide range instruments or connects the leak a narrow range instrument, D is incorrect since the failure results in a false low level signal and is not associated with narrow range instruments. Plausible if the examinee believes this leak affects narrow range or associates the Level 8 trip with wide range.

Reference Information:

SOP 23.601 (Pg 11-12) Instrument trip sheets FOS M-5701-2 Instrument location.

Plant Procedures 23.601 NUREG 1123 KA Catalog Rev. 2 223002 PCIS/NSSS 223002 K6. Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF:

223002 K6.04 Nuclear boiler instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILO 2015 Exam ILO 2017 Audit ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

Reactor Pressure Vessel Instrumentation Cognitive Terminal In accordance with approved plant procedures/references, given various controls and indications for operation of the RPV Instrumentation System: Describe the system response to loss of electrical power, various component or equipment failures, and abnormal operating conditions, as applicable.

ILT 2020 Final Version Page: 129 of 259 Question 51 Approved View

52 K/A Importance: 2.6 / 2.9 Points: 1.00 R52 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED 89849 The plant is operating at 100% power with the following:

Division 1 130V ESF Batteries are undergoing an equalizing charge.

T4100-B033, Battery Room A/C Unit is tripped.

8D22, Aux Bldg Battery Room A/C Unit Trouble is alarming.

NO AIR FLOW ACROSS BATT ROOM A/C UNIT white light is ON on H11-P808.

Division 1 Battery Room Temperature is 72°F.

T4100-C007 and T4100-C008, Division 1 Battery Room East and West Exhaust Fans are in AUTO.

What is the status of T4100-C007 and T4100-C008, Division 1 Battery Room East and West Exhaust Fans?

A. Running to prevent unacceptable room temperatures.

B. Running to prevent undesirable buildup of explosive hydrogen gas.

C. Will run if Battery Room ambient temperature exceeds 75°F to prevent unacceptable room temperatures.

D. Will NOT run with the Battery Room AC Unit tripped due to lack of cooling water flow through the cooling coils.

Answer: B ILT 2020 Final Version Page: 130 of 259 Question 52 Approved View

Answer Explanation:

NOTE: This question was modified by changing the stem such that a previously incorrect distractor is correct.

From 23.426, Reactor Building Heating Ventilation and Air Conditioning:

Air Conditioner (T4100-B033) fan units will operate continuously, but compressor will cycle on and off to supply cool air to Battery Room when Essential Battery Room ambient temperature exceeds 75°F. In AUTO, Exhaust Fans (T4100-C007 and T4100-C008 for Division 1) will auto start if air flow stops, preventing undesirable buildup of explosive hydrogen gas in the Battery Room(s).

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor would be true if the battery room exhaust fans auto started, to protect the battery or battery components from excessive temperature, when the B033 Battery Room AC Unit tripped off.

This distractor is incorrect because the battery room exhaust fans only auto start upon receipt of a no air flow condition to prevent undesirable buildup of explosive hydrogen gas.

C. This distractor would be true if the battery room exhaust fans started, to draw air across the cooling coils and cool the battery rooms, when ambient temperature reached 75F, which is plausible because this is the setpoint at which the fans associated with the B033, Battery Room AC Unit, auto start to prevent excessive room temperatures. This distractor is incorrect because the exhaust fans will be running due to low B033 air flow and do not auto start on high temperature in the Battery Rooms.

D. This distractor would be true if an interlock existed between the battery room exhaust fan auto start and the status of the B033 unit, which is plausible if the candidate incorrectly recalled that the fans serve to cool the room and work in conjunction with the B033 unit. In other words, if the candidate concluded that the exhaust fans draw air across cooling coils and thus are needed to cool the room, then this answer would be selected. This answer is incorrect because the AC fans operate continuously (for ventilation) and the compressor cycles (for cooling) based on room temperature, but the exhaust fans will only auto start upon receipt of a no air flow condition from the battery room AC unit to protect against hydrogen buildup.

Reference Information:

23.426, Reactor Building Heating Ventilation and Air Conditioning.

8D22, Aux Bldg Battery Room A/C Unit Trouble.

ILT 2020 Final Version Page: 131 of 259 Question 52 Approved View

Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 263000 DC Electrical Distribution 263000 K5 Knowledge of the operational implications of the following concepts as they apply to D.C.

ELECTRICAL DISTRIBUTION :

263000 K5.01 2.6/2.9 Hydrogen generation during battery charging 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level Modified RO Associated objective(s):

DC Electrical Distribution Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: Describe the impact of a loss of Battery Room ventilation (air flow and/or cooling) while charging a battery.

ILT 2020 Final Version Page: 132 of 259 Question 52 Approved View

53 K/A Importance: 2.8 Points: 1.00 R53 Difficulty: 3.00 Level of Knowledge: High Source: NEW 92277 The plant is at 100% power when 5D14, TBCCW Head Tank Level Hi/Low is received.

When dispatched, the TB Rounds NO reports:

P43-R401, TBCCW Head Tank Level Indicator, is reading +8.

P43-F400, TBCCW Head Tank Demin Water Makeup LCV, is open.

(1) What is the impact?

(2) What action should be taken?

A. (1) TBCCW Head Tank Level is high.

(2) Drain the Head Tank to Radwaste.

B. (1) The TBCCW Pumps should have tripped.

(2) Place the MODE Switch in Shutdown.

C. (1) The TBCCW Head Tank could overflow.

(2) Close P4300-F001, TBCCW Head Tank Demin Water Makeup LCV Inlet Iso Vlv.

D. (1) If TBCCW Head Tank Level drops 2 inches, the TBCCW Pumps will Trip.

(2) Slowly throttle open P4300-F003, TBCCW Head Tank Demin Water Makeup LCV Bypass Vlv.

Answer: C ILT 2020 Final Version Page: 133 of 259 Question 53 Approved View

Answer Explanation:

Note: TBCCW Head Tank Level indications are a common misconception among operator candidates because of legacy issues with old level indicators and information in references that can be confusing to new operators. For example, 5D14, TBCCW Head Tank Level Hi/Low comes in at 17 decreasing in the head tank, which corresponds to -7 on the local indicator, P43-R401. 5D14, TBCCW Head Tank Level Hi/Low also alarms at 32 increasing in the head tank, which corresponds to +8 on the local indicator, P43-R401.

Per ARP 5D14, and the report of indicated level from the field, the examinee should determine that (1)

Head Tank level is high and the makeup valve has failed open. (2) The examinee should determine that the makeup LCV should be isolated.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. TBCCW Head Tank level is high so the examinee could determine that the correct action is to drain the tank, which is true. However, TBCCW needs to be drained to drums, instead of Radwaste, due to corrosion inhibitor in the water.

B. As stated above, TBCCW Head Tank Level indication is confusing and the examinee could determine that the TBCCW Pumps should have tripped requiring the mode switch to be placed in Shutdown per 20.128.01, Loss of TBCCW AOP, which includes an override that requires this action for a total loss of TBCCW. This distractor is incorrect because Head Tank Level is high, thus the pumps should not have tripped.

D. Again, TBCCW Head Tank Level indication is confusing and the examinee could determine that the indication provided is 8 inches above the bottom of the tank, which means the pumps will trip if level drops 2 inches, since the TBCCW pumps trip at 6 inches in the head tank per 23.128, TBCCW SOP, P&L 3.2.1. If tank level was 2 above the pump trip, the correct action, per ARP 5D14, is to add water to the tank. This distractor is incorrect because the indication given (+8 on the P43-R401) is actually 32 above the bottom of the tank. If level dropped 2 inches, it would still be 30 above tank bottom, well above the pump trip.

Reference Information:

5D14, TBCCW Head Tank Level Hi/Low.

23.128 TBCCW SOP.

20.128.01, Loss of TBCCW AOP.

NUREG 1123 KA Catalog Rev. 2 400000 Component Cooling Water System 400000 A2. Ability to (a) predict the impacts of the following on the CCWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation:

400000 A2.02 2.8/3 High/low surge tank level 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO ILT 2020 Final Version Page: 134 of 259 Question 53 Approved View

Associated objective(s):

Turbine Building Closed Cooling Water Cognitive Terminal In accordance with approved plant procedures/references, given various controls and indications for operation of the TBCCW System: Discuss effective monitoring and control of the system using local and remote controls, indications, computer displays, alarms, and data-logging devices.

ILT 2020 Final Version Page: 135 of 259 Question 53 Approved View

54 K/A Importance: 2.9/2.9 Points: 1.00 R54 Difficulty: 3.00 Level of Knowledge: High Source: NEW 90487 The plant is operating at 100% power with the following:

C11-F002A, A CRD Flow Control Valve, is in service at about 40% open.

C11-K612, CRD Flow Controller is in MANUAL for troubleshooting excessive noise in the control circuit.

C1152-F003, CRD Drive/Cooling Water Pressure Control Valve is approximately 50% open.

C11-R800, CRD Hydraulic Flow Indicator, reads 63 gpm, steady.

No rod motion is in progress.

If a Reactor Scram were to occur:

(1) What action must be taken?

(2) If this action is taken, how will Cooling Water Header Pressure respond?

A. (1) Manually close C11-F002A, CRD Flow Control Valve.

(2) Lower to 0 psi.

B. (1) Manually close C11-F002A, CRD Flow Control Valve.

(2) Lower but remain above 0 psi.

C. (1) Manually close C1152-F003, CRD Drive/Cooling Water Pressure Control Valve.

(2) Lower to 0 psi.

D. (1) Manually close C1152-F003, CRD Drive/Cooling Water Pressure Control Valve.

(2) Lower but remain above 0 psi.

Answer: B ILT 2020 Final Version Page: 136 of 259 Question 54 Approved View

Answer Explanation:

23.106, Control Rod Drive Hydraulic System, Section 6.4 CRD Flow Stabilizing / CRD Flow Control Valves Operation, Page 46 contains a NOTE that states "If Reactor Scram occurs while operating in manual, the FCV must be closed from the controlling station."

With the CRD Flow Controller in Manual as stated in the stem, the candidate must recall that the CRD FCV must be manually closed in the event of a Reactor Scram.

The candidate must then recall the system drawing (FOS M-5703-1) and the location of the FCV in the system. Additionally, the candidate must recall that the FCV has a mechanical block that prevents the FCV from fulling closing, thus passing approximately 15 gpm to maintain cooling water to the CRDMs following a scram. Because of this, the candidate must recall that Cooling Water Header Pressure will lower, due to closing the FCV, but will not lower to 0 psi since the FCV is still partially open and the PCV is 50% open.

A is incorrect because, although the FCV must be closed in the event of a Reactor Scram, the mechanical block on the valve prevents downstream pressure (i.e., pressure on the Cooling Water Header) from going to 0 psi.

C is incorrect because the procedure directs manually closing the FCV upon a scram, not the PCV. This is plausible because closing the PCV would perform the same function (limit system flow to prevent the CRDH pump from going into runout on a scram), however it is not procedurally directed nor permitted. IF the candidate determined that closing the PCV was the correct action, Cooling Water Header Pressure going to 0 psi is plausible because doing so would block flow to the Cooling Water Header.

D is incorrect because the procedure directs manually closing the FCV upon a scram, not the PCV. This is plausible because closing the PCV would perform the same function (limit system flow to prevent the CRDH pump from going into runout on a scram), however it is not procedurally directed nor permitted. If the candidate recalled that cooling water flow does not lower to 0 on a scram, he/she could conclude that flow bypasses the PCV (perhaps through the Stabilizing Valves) or that the PCV is notched to provide some Cooling Water Header Pressure even with the PCV closed.

References:

23.106, Control Rod Drive Hydraulic System, Section 6.4 CRD Flow Stabilizing / CRD Flow Control Valves Operation.

M-5703-1, CRD System FOS.

NUREG 1123 KA Catalog Rev. 2 201001 CRDH System 201001 A1. Ability to predict and/or monitor changes in parameters associated with operating the CONTROL ROD DRIVE HYDRAULIC SYSTEM controls including:

201001 A1.02 Ability to predict and/ or monitor changes in parameters associated with operating the CRD HYDRAULIC SYSTEM controls including: CRD cooling water header pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (6) Design, components, and function of reactivity control mechanisms and instrumentation.

10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO ILT 2020 Final Version Page: 137 of 259 Question 54 Approved View

Associated objective(s):

Control Rod Drive Hydraulics Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Explain Control Rod Drive Hydraulic system operations necessary to support implementation of emergency operating procedure actions outside the control room.

ILT 2020 Final Version Page: 138 of 259 Question 54 Approved View

55 K/A Importance: 3.5 Points: 1.00 R55 Difficulty: 3.00 Level of Knowledge: Higher Source: NEW 90287 cog level A reactor startup is in progress. Plant conditions are as follows:

Feed flow is 24% and stable.

Steam flow is 23% and stable.

If the P603 operator attempts to withdraw a control rod that is NOT in accordance with the approved withdrawal sequence, the Rod Worth Minimizer ___(1)___ allow the rod to be withdrawn because reactor power is ___(2)___ the RWM Low Power Setpoint.

A. (1) will NOT (2) above B. (1) will (2) above C. (1) will NOT (2) below D. (1) will (2) below Answer: B ILT 2020 Final Version Page: 139 of 259 Question 55 Approved View

Answer Explanation:

Per 23.608, The RWM enforces adherence to the selected (computer program) control rod sequence up to the Low Power Setpoint (LPSP). RWM Sequence Enforcement restricts movement of Control Rods that are not in compliance with the rod by rod order selected as listed in a programmed sequence below the LPSP. No rod motion that would result in an insert or withdraw error is permitted. The LPSP is determined by the digital outputs of the Feedwater Control System (FWCS), which are driven by flow transmitters used for steam and feedwater flow input to the FWCS. When increasing power, LPSP enforcement is in effect until total feedwater flow is > 12.275% and total steam flow is > 12.25%.

In the Transition zone, between LPSP and LPAP (Low Power Alarm Point), the RWM does not enforce the operating sequence, and is automatically BYPASSED with respect to enforcement of rod movement.

In this instance power is rising, and is ABOVE the LPSP, below the LPAP, and is in the "transition zone".

Therefore a rod withdrawal error will NOT be blocked by the RWM Distractor Explanation:

A is incorrect because rod blocks are not enforced by the RWM when above the LPSP. This is plausible since the RBM system enforces rod blocks when power is above its setpoint and applicants may choose this if they confuse RWM and RBM automatic actions relative to reactor power.

C is incorrect because power is currently above the LPSP, and rod blocks are not enforced by the RWM when above the LPSP. This is plausible if applicants confuse the LPSP and LPAP, believe that the RWM is in the enforcement zone.

D is incorrect because the RWM is above the LPSP, not below. This is plausible if applicants confuse the LPSP and LPAP, and incorrectly believe that the RWM will not enforce blocks when below the LPSP.

Reference Information:

Per 23.608, Rod Worth Minimizer LP-OP-315-0013 K/A Match Justification:

This question matches the selected k/a because RO applicants must use plant parameters to monitor proper operation of the RWM.

NUREG 1123 KA Catalog Rev. 2 201006 RWM System 201006 A3. Ability to monitor automatic operations of the RODWORTH MINIMIZER SYSTEM (RWM)

(PLANT SPECIFIC) including:

201006 A3.04 3.5/3.4 Control rod movement blocks: P-Spec(Not-BWR6) 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO ILT 2020 Final Version Page: 140 of 259 Question 55 Approved View

Associated objective(s):

Rod Worth Minimizer Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: List the automatic features of Rod Worth Minimizer system operations.

ILT 2020 Final Version Page: 141 of 259 Question 55 Approved View

56 K/A Importance: 3.0/3.1 Points: 1.00 R56 Difficulty: 2.00 Level of Knowledge: Fund Source: BANK: FERMI 2001 90291 RO RETAKE NRC The reactor is operating at 100% power. A Reactor Engineer is performing a TIP run per 23.606 "TIP System"

C TIP detector is fully inserted into the core.

A small steam leak occurs inside the Drywell.

Drywell pressure is 1.75 psig.

How will the TIP system respond?

A. The shear valve will automatically fire immediately.

B. The ball valve will automatically close ONLY after operators withdraw the detector.

C. The ball valve will close after the detector automatically retracts into the chamber shield.

D. The shear valve will automatically fire after the detector automatically retracts into the chamber shield.

Answer: C ILT 2020 Final Version Page: 142 of 259 Question 56 Approved View

Answer Explanation:

Per TS 3.3.6.1 table 3.3.6.1-1, TIP isolation occurs on High Drywell Pressure, OR RPV Level 3.

Per 23.606:

"If the MAN VALVE CONTROL Switch is not in OPEN, the Ball Valve will automatically close when the detector is withdrawn into its storage position in the Chamber Shield. If the detector is inserted past the Chamber Shield Limit Switch (actuation of the Limit Switch opens the Ball Valve), and the Containment Isolation Circuit in the Valve Control Monitor is activated, the system automatically reverts to the Manual Reverse Mode of operation. The detector will be retracted to the Chamber Shield and the Ball Valve closed.

The Shear Valve is provided for emergency use only. If containment integrity is required (regardless of whether or not an actual tube leak has occurred) and for some reason the Detector cable cannot, or should not be withdrawn, or if the Ball Valve fails to close, the Shear Valve can be detonated. This cuts the tube and the cable, sealing off the Reactor side of the Guide Tube. The Shear Valve is actuated by a Keylock Switch located on the Valve Control Monitor. The Keylock Switch is designed to prevent accidental firing of the Shear Valve. Key may be obtained from the SM."

Therefore, in this instance, an isolation signal is present causing the TIP detector to automatically withdraw into the shield chamber, allowing the ball valve to auto close.

A is incorrect since the shear valve does not auto close. Plausible if applicants confuse operation of the ball and shear valve.

B is incorrect since the detector will auto retract prior to ball valve closure. Plausible if applicants believe manual retraction is necessary to allow isolation.

D is incorrect since the shear valve does not auto close. Plausible if applicants confuse operation of the ball and shear valve.

This question meets the selected K/A because RO applicants must demonstrate knowledge required to monitor automatic operation of isolation valves in the TIP system.

NUREG 1123 KA Catalog Rev. 2 215001 Traversing In-Core Probe 215001 A4. Ability to monitor automatic operations of the TRAVERSING IN-CORE PROBE including:

215001 A4.03 3/3.1 Isolation valves: Mark-I&II(Not-BWR1) 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILO 2001 Retake Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental RO ILT 2020 Final Version Page: 143 of 259 Question 56 Approved View

Associated objective(s):

Traversing In-Core Probe Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: Describe Traversing In-Core Probe System automatic features.

ILT 2020 Final Version Page: 144 of 259 Question 56 Approved View

57 K/A Importance: 4.2/4.2 Points: 1.00 R57 Difficulty: 3.00 Level of Knowledge: higher Source: NEW 90507 The plant is in MODE 1 conducting a reactor startup.

Reactor power is 35% and stable.

The P603 operator selects and withdraws a central control rod per the approved pull sheet.

As the rod is being withdrawn, RBM A flux indication rises to 114% indicated flux.

The following annunciators are then received:

3D109, RBM UPSCALE/INOP.

3D113, CONTROL ROD WITHDRAWAL BLOCK.

These alarms are...

A. NOT VALID because RBMs are bypassed above 27% power.

B. VALID because RBMs are no longer bypassed above 27% power and the rod block setpoint is 107.2%.

C. VALID because RBMs are no longer bypassed above 27% power and the rod block setpoint is 112.2%.

D. NOT VALID because, although RBMs are no longer bypassed above 27%

power, the rod block setpoint is 117%.

Answer: D ILT 2020 Final Version Page: 145 of 259 Question 57 Approved View

Answer Explanation:

Per ARP 3D109, RBM UPSCALE/INOP, the setpoint for rod block initiation is 117.0% when reactor power is between 27% and 62%. With power at 35%, the RBM function is active, however the indicated flux in this instance has not exceeded the setpoint to generate a rod block. The RBM rod block function is not active below 27% power.

Therefore, the upscale and rod block annunciators are not valid.

Distractor explanation:

A is incorrect since the RBM is bypassed when power is less than 27%. Plausible if the applicants confuse the bypass functions of the RBM and the RWM, which is bypassed as power rises above a certain setpoint, opposite that of the RBM.

B is incorrect since the setpoint at 35% power is 117.0, not 107.2, and the alarms should not be received.

This is plausible since 107.2% is a valid setpoint for Rod Blocks but is true only when power is greater than 82%

C is incorrect since the setpoint at 35% power is 117.0, not 112.2, and the alarms should not be received.

This is plausible since 112.2% is a valid setpoint for Rod Blocks but is true only when power is between 62% and 82% power.

References:

23.607, Rod Block Monitor System procedure

3D109, RBM UPSCALE/INOP alarm response procedure This question matches the selected K/A because RO candidates must determine the validity of RBM related alarms based on plant conditions.

NUREG 1123 KA Catalog Rev. 2 215002 RBM System G2.4.46 Ability to verify that the alarms are consistent with the plant conditions 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO Associated objective(s):

Power Range Monitoring and Rod Block Monitoring Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: Discuss effective monitoring of the Power Range Neutron Monitoring System using local, remote, computer displays and alarms.

ILT 2020 Final Version Page: 146 of 259 Question 57 Approved View

58 K/A Importance: 3.4 Points: 1.00 R58 Difficulty: 3.00 Level of Knowledge: High Source: NEW 92561 A LOCA coincident with a Loss of Offsite Power has occurred.

EDGs 12 and 13 have failed to start.

RHR is running as follows:

RHR Pump A is running in Torus Cooling at 13,000 gpm.

RHR Pump D is running for LPCI injection at 13,000 gpm.

E1150-F010, RHR Crosstie Valve is closed.

The CRS wants to place Torus Spray in service and is asking for your recommendation.

Which Division(s) of RHR, if any, is(are) acceptable for use in the Torus Spray mode?

A. Division 1 ONLY.

B. Division 2 ONLY.

C. Either is acceptable.

D. Neither under the current conditions.

Answer: A ILT 2020 Final Version Page: 147 of 259 Question 58 Approved View

Answer Explanation:

Per 23.205, RHR System SOP, Section 9.6, Torus Spray Mode, prerequisite 9.6.1.2 requires RHR to be operating in either Torus Cooling or LPCI Modes prior to placing RHR in Torus Spray.

P&L 3.1.5 states that Flow of 14,000 gpm per RHR Pump must not be exceeded to avoid pump runout.

RHR pump D coupling limits its flow to 13,000 gpm.

RHR Pumps A (EDG 11) and C (EDG 12) supply Division 1 RHR and Pumps B (EDG 13) and D (EDG

14) supply Division 2 RHR. Because of the failure of EDGs 12 and 13, no additional pumps are available to support Torus Spray.

Torus Spray results in approximately 500 gpm of flow per step 9.6.2.5.b.

Therefore, the examinee must apply all the information above, and the flow indications provided in the stem, to recommend that Division 1 RHR be placed in Torus Spray.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. Division 2 RHR is running in the LPCI mode, which is acceptable per prerequisite 9.6.1.2 to be placed in Torus Spray. The examinee could fail to recall the RHR Pumps in each division or the examinee could fail to remember exactly which RHR Pump is limited to 13,000 gpm. Either of these could lead the examinee to determine that Division 2 RHR is preferred over Division 1. This distractor is incorrect because RHR Pump A is in Division 1 and RHR Pump D is at its flow limit.

C. Division 2 RHR is running in the LPCI mode, which is acceptable per prerequisite 9.6.1.2 to be placed in Torus Spray. The examinee could fail to recall that RHR Pump D has coupling issues that restrict its operation to 13,000 gpm. This distractor is incorrect because RHR Pump D is at its flow limit.

D. The examinee could incorrectly recall that the 13,000 gpm flow limit for RHR Pump D applies to all RHR Pumps and determine that neither division of RHR could be placed in Torus Spray, unless another pump was available, or Torus Cooling flow throttled. This distractor is incorrect because RHR Pump A has capacity to its flow limit of 14,000 gpm that would support Torus Spray.

Reference Information:

23.205, RHR System SOP.

NUREG 1123 KA Catalog Rev. 2 230000 RHR/LPCI: Torus/Suppression Pool Spray Mode 230000 K1. Knowledge of the physical connections and/or cause effect relationships between RHR/LPCI:TORUS/SUPPRESSION POOL SPRAY MODE and the following:

230000 K1.04 3.4/3.6 LPCI/RHR pumps 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO ILT 2020 Final Version Page: 148 of 259 Question 58 Approved View

Associated objective(s):

Residual Heat Removal Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: State major precautions and limitations, and major safety considerations for the Residual Heat Removal System, and describe their bases.

ILT 2020 Final Version Page: 149 of 259 Question 58 Approved View

59 K/A Importance: 2.9 Points: 1.00 R59 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED 92109 The plant is in MODE 5 with the following:

A combination of new and irradiated fuel is seated within the Reactor Pressure Vessel (RPV).

Spent Fuel and Control Rod blades are being shuffled in the RPV.

An irradiated fuel assembly has just been loaded on the Refuel Bridge main hoist and raised to the full up position.

A rupture of the return line from the Fuel Pool Cooling and Cleanup (FPCCU) pumps occurs.

Which of the following describes (1) the result of this break and (2) what is required by Technical Specifications?

(1) RPV water level lowers...

(2) Immediately suspend movement of...

A. (1) and partially uncovers the fuel assembly (2) fuel assemblies ONLY B. (1) but maintains the fuel assembly fully covered (2) fuel assemblies ONLY C. (1) and partially uncovers the fuel assembly (2) fuel assemblies AND control rods D. (1) but maintains the fuel assembly fully covered (2) fuel assemblies AND control rods Answer: D ILT 2020 Final Version Page: 150 of 259 Question 59 Approved View

Answer Explanation:

Note: This question was modified from Question 29 of the 2019 Nine Mile NRC Exam. It was modified by changing the part (2) of the question to make the question a better fit for the K/A.

Vacuum Relief Lines prevent siphoning of the Spent Fuel Storage Pool. Open-ended vent lines on each diffuser line extend downward below the surface of the water in the pool, when at normal level. In the event of a line break at any point upstream of the vacuum reliefs, pool level will lower due to siphoning, until the open ends of the vents below the water are exposed to the atmosphere. At this point, the vacuum driving the active siphoning will be broken, and the siphoning action will stop. However, Fuel Pool water will continue to seek its own level and gravity drain out through the diffuser line break until Fuel Pool level has equalized with the water level inside the diffuser line at the elevation of the bottom edge of the pipe at its highest point. Failure of the return line will result in water level dropping to 6823 (bottom of return piping). The candidate must recall that this still provides over 4 of water above a fuel assembly suspended by the refuel bridge (top of suspended fuel assembly is at 67710). Therefore, (1) the fuel assembly will still be fully covered.

However, normal level during refueling is with water at 6836, which provides 210 of water over the RPV flange (elevation 6626). Water level dropping to the bottom of the return piping (6823) means that there is 199 of water above the RPV flange, which is below the LCO 3.9.6 value of 206 above the RPV flange during movement of irradiated fuel assemblies within the RPV or during movement of new fuel assemblies or handling of control rods within the RPV, when irradiated fuel assemblies are seated within the RPV. Therefore, (2) per LCO 3.9.6, with RPV water level not within limit, the required action is to immediately suspend movement of fuel assemblies AND handling of control rods within the RPV.

Distractor Explanation:

Distractors are incorrect and plausible because:

(1) is plausible if the examinee thought that the fuel assemblies were higher when being transported by the refueling bridge or if the examinee failed to recall the existence of the vacuum relief lines and thought the water level would drop lower than the elevation of the top of suspended fuel. Both are incorrect because water level stopping at 6823, due to the vacuum relief function, still maintains the water coverage over suspended fuel as described above.

(2) is plausible if the examinee failed to recall the wording of LCO 3.9.6 and thought that it is only applicable to the movement of irradiated fuel in the RPV. This is incorrect because the second applicability statement indicates the LCO is also applicable to handling of control rods in the RPV, therefore Condition A requires suspending movement of fuel assemblies AND control rods Immediately (making it RO knowledge).

Reference Information:

LCO 3.9.6, RPV Water Level (Refueling Operations).

FPCCU Student Text (ST-OP-315-0015-001).

ILT 2020 Final Version Page: 151 of 259 Question 59 Approved View

NUREG 1123 KA Catalog Rev. 2 233000 Fuel Pool Cooling and Cleanup 233000 K3. Knowledge of the effect that a loss or malfunction of the FUEL POOL COOLING AND CLEAN-UP will have on the following:

233000 K3.08 2.9/3.5 Refueling operations 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level Modified RO Associated objective(s):

Fuel Pool Cooling and Cleanup Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: Identify Fuel Pool Cooling and Cleanup System related technical specifications, with emphasis on action statements requiring prompt actions (for example, one hour or less).

ILT 2020 Final Version Page: 152 of 259 Question 59 Approved View

60 K/A Importance: 2.9/3.0 Points: 1.00 R60 Difficulty: 2.00 Level of Knowledge: High Source: NEW 90367 The plant is starting up, after a refueling outage, with the following conditions:

MODE 1

Reactor Power ~ 20%

Bypass Valves open ~25%.

Main Turbine rolling at 1800 rpm for turbine trip testing.

IF the governor interlock switch is placed in EXERCISE PERMIT, turbine valve response will cause RPV Pressure to __(1)__ and RPV Level to __(2)__.

A. (1) Rise.

(2) Rise.

B. (1) Rise.

(2) Lower.

C. (1) Lower.

(2) Rise.

D. (1) Lower.

(2) Lower.

Answer: C ILT 2020 Final Version Page: 153 of 259 Question 60 Approved View

Answer Explanation:

Turbine Generator off load valve gear stroke and trip testing tests Main Turbine valves when the turbine is shutdown by resetting the turbine trip and placing the governor in the exercise permit mode to allow cycling of HP Stop, HP Control, and LP Intercept valves (valves that normally respond during a turbine trip). However, if Exercise Permit is selected with pressure in the 52-inch manifold, the regulator setpoint is biased 30 psig lower. The bypass valves are then no longer be controlling 52-inch manifold pressure at 944 psig (nominally). The setpoint to the regulator is approximately 914 psig, and bypass valve demand is greater than 100%. The reactor rapidly depressurizes and reactor level swells above Level 8, Note: This concept is operationally significant due to an event that occurred in 2008 whereby this switch was taken to Exercise Permit causing an RPV Pressure and Level transient. Because of this event, Precaution and Limitation 3.2.1 was added to 23.109, to warn the operator against this action. The P&L states DO NOT place the Governor Interlock switch in EXERCISE PERMIT with any pressure in the 52" Manifold. This will cause the regulators' setpoint to be lowered approximately 30 psig and the Bypass Valve to fully open. An RPV water level and pressure transient will result.

This question tests the candidates ability to recall this caution and conclude that RPV pressure will lower and RPV level will rise if the switch in the stem is taken to Exercise Permit under the stated plant conditions.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. (1) The candidate could incorrectly recall that placing the switch in Exercise Permit will cause the turbine valves to rapidly close causing RPV pressure to rise, which is incorrect as described above.

(2) This part is correct.

B. (1) The candidate could incorrectly recall that placing the switch in Exercise Permit will cause the turbine valves to rapidly close causing RPV pressure to rise, which is incorrect as described above.

(2) The candidate could logically determine that, since he/she thinks RPV pressure will rise per part (1),

the pressure rise will cause collapsing of steam bubbles in the core region and lowering of RPV level in the downcomer region, due to shrink. However, since part (1) is incorrect, this part is incorrect because swell will occur instead of shrink.

D. (1) Correct.

(2) The candidate could conclude that RPV level will lower because the candidate could determine that the rapid valve opening will cause shrink to occur in the core (fuel) region due to the pressure reduction, which would draw water from the downcomer region, where RPV water level is sensed, which would cause RPV level to lower. This is incorrect because the pressure reduction is felt across the entire RPV, not just the core (fuel) region, and it will cause some water in the RPV (at saturated conditions) to expand and form steam bubbles, thus causing indicated level to "swell" in the downcomer.

Reference Information:

23.109, Turbine Operating Procedure.

ILT 2020 Final Version Page: 154 of 259 Question 60 Approved View

NUREG 1123 KA Catalog Rev. 2 241000 Reactor/Turbine Pressure Regulating System 241000 K4. Knowledge of REACTOR/TURBINE PRESSURE REGULATING SYSTEM design feature(s) and/or interlocks which provide for the following:

241000 K4.13 2.9/3.0 Turbine trip testing: Plant-Specific 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New RO Associated objective(s):

Governor/Pressure Control Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: State major precautions and limitations, and major safety considerations for the Governor/Pressure Control System, and describe their bases.

ILT 2020 Final Version Page: 155 of 259 Question 60 Approved View

61 K/A Importance: 2.8/3.1 Points: 1.00 R61 Difficulty: 3.00 Level of Knowledge: High Source: MODIFIED 90368 While operating at 92% Power, alarms were received on the H11-P804 panel. Among them is 4D102, Turbine Valve Position Abnormal.

Upon investigation, the CRLNO reports that the following valves are CLOSED:

N3021-F013C, #3 Low Pressure Intercept Valve (LPIV) from the East MSR to the Center LP Turbine

N3021-F013E, #5 Low Pressure Intercept Valve (LPIV) from the East MSR to the South LP Turbine.

Plant power level increased to 93%.

Which of the following actions is required?

A. Reduce Reactor Power to <65%.

B. Reduce Reactor Power to <91.5%.

C. Scram the reactor and trip the main turbine.

D. Maintain the remaining open LP Intercept Valves <82% open.

Answer: A ILT 2020 Final Version Page: 156 of 259 Question 61 Approved View

Answer Explanation:

Note: This question was modified from a question previously used on an NRC initial license exam by modifying conditions in the stem such that a previously incorrect distractor (A) is now correct and the previously correct answer (C) is now incorrect.

Per 23.109 P&L 3.13.1 two outlets from an MSR shell shall not be closed at loads greater than 64% and per ARP 4D102, Turbine Valve Position Abnormal, Closure of LP steam inlets to two LP Turbines by closed Stop or Intercept Valves (two outlets from one MSR) requires the operator to immediately reduce Reactor Power to < 65%.

Therefore, the examinee must first evaluate the valve lineup given in the stem of the question, then apply that information to the turbine operating limits above, and finally determine that load must be reduced to

<65%.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. Closure of a Turbine Control Valve requires reactor power be lowered to 91.5% or lower in accordance with 23.109, P&L 3.12.3. This distractor is incorrect because the valve lineup given requires power be lowered <65%.

C. Note: This was the previously correct distractor. This distractor is plausible because 23.109 P&L 3.12.2 and ARP 4D102, Turbine Valve Position Abnormal, do not allow turbine operation if both steam inlet lines to any one LP Turbine are isolated by a closed Stop or Intercept Valve, thus requiring the turbine to be tripped and the reactor scrammed. This distractor is incorrect because the valve lineup given in the stem indicate that 2 inlets from the same MSR have closed, not 2 inlets to the same LP turbine (which would be from opposite MSRs).

D. Closure of a Turbine Control Valve requires reactor power be lowered to 91.5% or lower in accordance with 23.109, P&L 3.12.3. This distractor is incorrect because the valve lineup given requires power be lowered <65%.

Reference Information:

4D102, Turbine Valve Position Abnormal.

23.109, Turbine Operating Procedure.

Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 245000 Main Turbine Generator and Auxiliary System 245000 K5. Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS:

245000 K5.02 2.8/3.1 Turbine operation and limitations 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level Modified RO ILT 2020 Final Version Page: 157 of 259 Question 61 Approved View

Associated objective(s):

Turbine Steam Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: Describe Turbine Steam System precautions and limitations.

ILT 2020 Final Version Page: 158 of 259 Question 61 Approved View

62 K/A Importance: 2.7/2.8 Points: 1.00 R62 Difficulty: 2.00 Level of Knowledge: Fund Source: NEW 90747 Which of the following Condensate System pumps are impacted by a loss of 4160VAC bus 64A?

A. Center Condenser Pump ONLY.

B. North AND South Condenser Pumps ONLY.

C. Center Condenser Pump, AND the Center Heater Feed Pump.

D. North and South Condenser Pumps, AND the East and West Heater Feed Pumps.

Answer: C ILT 2020 Final Version Page: 159 of 259 Question 62 Approved View

Answer Explanation:

Per AOP 20.300.64A attachment 1, the Center Condenser pump, and the Center HFP are deenergized upon loss of bus 64A.

A is incorrect since the Center Heater Feed pump is lost also. Plausible if applicants cannot correctly recall major condensate system component power supplies.

B is incorrect since these pumps are powered by bus 65D. Plausible if applicants cannot correctly recall major condensate system component power supplies.

D is incorrect since these pumps are powered by bus 65D. Plausible if applicants cannot correctly recall major condensate system component power supplies.

References:

AOP 20.300.64A, Loss of 64A AOP 20.300.65D, Loss of 65D This question matches the selected K/A because RO Applicants must correctly recall the power supplies to major Condensate system pumps.

NUREG 1123 KA Catalog Rev. 2 256000 Reactor Condensate System 256000 K2. Knowledge of electrical power supplies to the following:

256000 K2.01 System pumps 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (3) Mechanical components and design features of reactor primary system.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

Condensate Cognitive Terminal In accordance with approved plant procedures/references, under all conditions of the Condensate System: Identify the normal and alternate power supplies to system components.

ILT 2020 Final Version Page: 160 of 259 Question 62 Approved View

63 K/A Importance: 3.3 Points: 1.00 R63 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 92588 A reactor startup is in progress with power at 2% and RPV pressure at 200 psig. The following alarms are received:

3D82, MN STM LINE CH A/B/C/D RADN MONITOR HI-HI.

3D83, MN STM LINE CH A/B/C/D RADN MONITOR UPSCALE.

3D20, 2 MINUTE HOLDUP PIPE RADN MONITOR UPSCALE.

3D24, 2 MINUTE HOLDUP PIPE RADN MONITOR UPSCALE TRIP.

You subsequently observe the following:

All Main Steam Isolation Valves (MSIVs) are open.

The East Mechanical Vacuum Pump (MVP) is running.

All Steam Jet Air Ejectors (SJAEs) are shutdown.

The Gland Seal Exhausters (GSEs) are shutdown.

Which of the following describes the action you will take and the result on Condenser Vacuum?

A. The MVP failed to trip. Trip the MVP. Condenser Vacuum will degrade.

B. A GSE should be running. Start a GSE. Condenser Vacuum will improve.

C. The MSIVs failed to close. Close the MSIVs. Condenser Vacuum will improve.

D. The SJAEs should not have tripped. Start a SJAE. Condenser Vacuum will improve.

Answer: A ILT 2020 Final Version Page: 161 of 259 Question 63 Approved View

Answer Explanation:

Per 3D82 Auto Actions, the examinee should recognize that the running MVP(s) should have tripped since power is <10% (the examinee must determine that power is <10% based on RPV Pressure being only 200 psig).

The examinee should determine that the correct action is the shut down the running MVP, which will result in degrading condenser vacuum.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The examinee could fail to recall that the running GSE will trip when 3D82 is received <10% power, which is plausible because prior to 2019 the GSEs did not trip when 3D82 was received This could lead the examinee to determine that one of the GSEs should be started, which would improve main condenser vacuum. This is incorrect because the running GSE should have tripped as described above and if the examinee attempted to start one, it would trip.

C. The examinee could determine that the MSIVs should have closed, which is plausible because prior to 2019 the MSIVs did close when 3D82 was received. This is incorrect because the MSIVs should NOT have automatically closed.

D. The examinee could fail to recall when the SJAEs are placed in service and therefore determine that the tripped, which is plausible because SJAEs at Fermi 2 do not trip on high radiation level. This could lead the examinee to determine that a SJAE should be started, which would improve condenser vacuum. This is incorrect because SJAEs are not placed in service until after 300 psig reactor pressure to ensure they have sufficient steam pressure to operate..

Reference Information:

3D82, MN STM LINE CH A/B/C/D RADN MONITOR HI-HI.

3D83, MN STM LINE CH A/B/C/D RADN MONITOR UPSCALE.

3D20, 2 MINUTE HOLDUP PIPE RADN MONITOR UPSCALE.

3D24, 2 MINUTE HOLDUP PIPE RADN MONITOR UPSCALE TRIP NUREG 1123 KA Catalog Rev. 2 271000 Offgas System 271000 A1. Ability to predict and/or monitor changes in parameters associated with operating the OFFGAS SYSTEM controls including:

271000 A1.01 3.3/3.2 Condenser vacuum 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (5) Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New NRC Early Review RO Associated objective(s):

Off Gas Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: Describe general Off Gas System operation, including component operating sequence, normal operating parameters, and expected system response.

ILT 2020 Final Version Page: 162 of 259 Question 63 Approved View

ILT 2020 Final Version Page: 163 of 259 Question 63 Approved View

64 K/A Importance: 2.8/3.0 Points: 1.00 R64 Difficulty: 2.00 Level of Knowledge: Fund Source: BANK: FERMI 2008 90509 NRC Loss of which of the following power sources will de-energize D11-K609A, Fuel Pool (EAST) Vent Exhaust Duct Radiation Monitor?

A. Modular Power Unit (MPU) 1 B. Modular Power Unit (MPU) 3 C. Reactor Protection System (RPS) A D. Reactor Protection System (RPS) B Answer: C ILT 2020 Final Version Page: 164 of 259 Question 64 Approved View

Answer Explanation:

Correct Answer: C RPS A provides power to D11-K609A, Fuel Pool (EAST) Vent Exhaust Duct Radiation Monitor.

Plausible Distractors:

A is plausible; MPU 1 provides power for CC HVAC Div 1 Radiation Monitor.

B is plausible; MPU 3 provides power for RB Exhaust Plenum SPING.

D is plausible; RPS B provides power to D11-K609B & D

References:

23.625, PROCESS GASEOUS RADIATION MONITORING NUREG 1123 KA Catalog Rev. 2 272000 Radiation Monitoring System 272000 K6. Knowledge of the effect that a loss or malfunction of the following will have on the RADIATION MONITORINGSYSTEM :

272000 K6.03 2.8/3 A.C. power 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (7) Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Fermi 2 NRC Exam Usage ILO 2008 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental RO Associated objective(s):

Process Radiation Monitoring System Cognitive Terminal In accordance with approved plant procedures/references, under all conditions of the Process Radiation Monitoring System: Identify the normal and alternate power supplies to system components.

ILT 2020 Final Version Page: 165 of 259 Question 64 Approved View

65 K/A Importance: 3.3/3.7 Points: 1.00 R65 - Difficulty: 3.00 Level of Knowledge: Low Source: NEW 97287 POST SUBMIT TAL VERSIO N

The plant is in MODE 1 at 100% power.

17D43, Rail Airlock Outer Door Seal Pressure Low is received.

Per the Alarm Response Procedure for 17D43:

You should verify that __ (1) __ NIAS pressure is normal on P50-R801, NIAS Header Pressure Recorder.

If Outer Door Seal Pressure is less than __ (2) __ psig, interlocks will prevent opening of the Inner Door to inspect the Outer Door seals and components from inside the airlock.

A. (1) Div 1 (2) 5 B. (1) Div 1 (2) 28 C. (1) Div 2 (2) 5 D. (1) Div 2 (2) 28 Answer: C ILT 2020 Final Version Page: 166 of 259 Question 65 Approved View

Answer Explanation:

The initial response for 17D43 is to (1) verify Div 2 NIAS pressure is normal on P50-R801, NIAS Header Pressure Recorder.

Per the note in 17D43, if seal pressure is (2) <5 psig, then the Inner Door will be prevented from opening.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. (1) is plausible because the inner airlock door seals are pressurized with Div 1 NIAS, so the candidate could select Division 1 if he/she confused which division of NIAS supplied which door's seals. This is incorrect because 17D43 requires checking Div 2 NIAS pressure and not Div 1. (2) is plausible because 5 psig is correct.

B. (1) is plausible because the inner airlock door seals are pressurized with Div 1 NIAS, so the candidate could select Division 1 if he/she confused which division of NIAS supplied which door's seals. This is incorrect because 17D43 requires checking Div 2 NIAS pressure and not Div 1. (2) is plausible because 17D42 alarms at 28 psig, so the candidate could relate the alarm setpoint with the interlock between the two doors. This is incorrect because the interlock occurs when door pressure drops <5 psig.

C. (1) is correct. (2) is plausible because 17D42 alarms at 28 psig, so the candidate could relate the alarm setpoint with the interlock between the two doors. This is incorrect because the interlock occurs when door pressure drops <5 psig.

Reference Information:

17D43, Rail Airlock Outer Door Seal Pressure Low.

23.428, Secondary Containment Airlocks and Penetrations.

NUREG 1123 KA Catalog Rev. 2 290001 Secondary Containment 290001 A2. Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT ;

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

290001 A2.01 3.3/3.7 Personnel airlock failure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b) (9) Shielding, isolation, and containment design features, including access limitations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New NRC Early Review RO Associated objective(s):

ILT 2020 Final Version Page: 167 of 259 Question 65 Approved View

66 K/A Importance: 4.3/4.6 Points: 1.00 R66 Difficulty: 3.00 Level of Knowledge: High Source: BANK (NOT USED 90148 ON NRC EXAM)

The plant is at 100% power.

You are performing the Core Thermal Limit Verification in accordance with 24.000.02. While reviewing the latest core monitor edit you observe the following:

Reactor Pressure .......... 1020 psig

Reactor Level .................. 197 inches

MCPR .......................1.27

MAPRAT ..................................................................................... 0.887

MFLPD ........................................................................................ 0.978

MFLCPR ..................................................................................... 0.997 What is the significance of this information and what action is required?

A. All limits are met; Reduce power to prevent exceeding an administrative limit.

B. Tech Spec Safety Limits are being exceeded; Insert all Control Rods within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

C. A Thermal Limit LCO is NOT being met; Restore the ratios to within limits within two hours.

D. Only an administrative thermal limit is NOT being met; Notify the Reactor Engineering Supervisor.

Answer: D ILT 2020 Final Version Page: 168 of 259 Question 66 Approved View

Answer Explanation:

Note: This question matches the K/A because performance of the 24.000.02, Core Thermal Limit Verification, is an administrative procedure used by licensed operators to evaluate for proper reactivity management. The K/A statement is knowledge of procedures, guidelines or limitations associated with Reactivity Management. Administrative limits are only listed in MOP19. The candidate must have specific knowledge of this procedure to determine that conditions in the stem of the question indicate that a limitation associated with Reactivity Management has been exceeded.

Upon review of the data, the operator should determine that all Tech Spec Thermal Limit SLs and LCOs are met, however the MFLCPR ratio is >0.990, which is above the Admin Limit of MOP 19. The operator should determine that the Supervisor, Reactor Engineering should be notified.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The operator could assume that, since all ratios are <1.0, then all limits are met. The operator could then determine that power reduction is necessary to avoid exceeding 1.0. This is incorrect since an administrative limit has already been exceeded.

B. The operator could incorrectly determine that a TS SL is not met, and then determine that the actions for a violated SL must be taken.

C. The operator could determine that a Thermal Limit LCO is not met, and then determine that the actions for not meeting one of these LCOs must be taken. This is incorrect since, if all ratios are <1.0, then all thermal limits are met.

Reference Information:

MOP19, Reactivity Management, Step 5.7.5 Tech Spec Thermal Limit LCOs 3.2.1, 3.2.2 and 3.2.3 Tech Spec Safety Limits Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 G2.1.37 Knowledge of procedures, guidelines, or limitations associated with reactivity management 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

Core and Fuel Cognitive Terminal In accordance with approved plant procedures, given various controls and indications for system operations: Discuss effective monitoring of the Core and Fuel using local, remote, computer displays and alarms.

ILT 2020 Final Version Page: 169 of 259 Question 66 Approved View

67 K/A Importance: 3.7/3.8 Points: 1.00 R67 Difficulty: 2.00 Level of Knowledge: High Source: NEW 90527 The plant was at 100% when an ATWS occurred.

The crew has entered 29.100.01 sheet 1, RPV Control.

ALL "ATWS actions" are complete and the crew has transitioned to 29.100.01 sheet 1A, RPV Control-ATWS.

RPV level and pressure are lowering.

Alarm 2D27, REACTOR LOW PRESSURE, is received.

The CRLNO responding to the alarm should report...

A. "Annunciator 2D27, Reactor Low Pressure - expected alarm".

B. "Reactor Low Pressure - low pressure systems are aligning for injection".

C. "Reactor Low Pressure - low pressure systems are terminated and prevented".

D. "Reactor Low Pressure - low pressure systems ARE NOT aligning for injection".

Answer: C ILT 2020 Final Version Page: 170 of 259 Question 67 Approved View

Answer Explanation:

The applicant must recall that, per EOP sheet 1, the "ATWS actions" include terminating and preventing injection to allow RPV level to lower.

Per ODE-10, Ops department expectations for EOP implementation:

"When 2D27 alarms or when reactor pressure is less than 461 psig, announce Reactor Pressure Low -

Low Pressure systems are (are not) aligning for injection If Terminate and Prevent has occurred, this should be modified to Reactor Pressure Low - Low Pressure systems are terminated and prevented.

Therefore, the operator must report RPV Pressure Low - Low Pressure systems are terminated and prevented.

A is incorrect and plausible because this is a typical report for an expected annunciator during non-EOP conditions, and applicants may believe that standard annunciator reports are applicable in this condition.

B is incorrect and plausible because this is the required report during non-ATWS EOP activity when low pressure systems automatically align for RPV injection at the low pressure setpoint, and applicants may not realize that terminate & prevent has been completed.

D is incorrect and plausible because this is the required report during non-ATWS EOP activity when low pressure systems have been prevented from aligning for auto injection, and applicants may not recall that there is a unique report required for terminate & prevent conditions.

References:

ODE-10, Emergency Operating Procedure Expectations 29.100.01 sheet 1, RPV Control This question matches the selected K/A since RO applicants must recall the station's requirements for verbal communications while implementing EOPs.

Plant Procedures ODE-10 EOP Expectations NUREG 1123 KA Catalog Rev. 2 G2.1.38 3.7/3.8 Knowledge of the station's requirements for verbal communications when implementing procedures 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New NRC Early Review RO ILT 2020 Final Version Page: 171 of 259 Question 67 Approved View

Associated objective(s):

Introduction to Emergency Operating Procedures Cognitive Terminal Given a set of plant parameters that meet the entry conditions, Explain the functions performed by the Shift manager, Control Room Supervisor, Nuclear Supervising Operators, Nuclear Operators, and Shift Technical Advisor during implementation of the Emergency Operating Procedures, in accordance with Fermi 2 Emergency Operating Procedures ILT 2020 Final Version Page: 172 of 259 Question 67 Approved View

68 K/A Importance: 3.7/4.1 Points: 1.00 R68 Difficulty: 3.00 Level of Knowledge: Low Source: BANK: FERMI 2017 90150 I & C technicians are preparing to perform a surveillance procedure on an instrument. They observe that a prerequisite to the procedure cannot be met under present plant conditions.

Omitting the prerequisite will not affect the performance of the procedure, and the surveillance will be past its critical date the next day.

Which of the following selections indicates who can waive the prerequisite and how this waiver must be documented after the prerequisite is marked as N/A in order to continue with the surveillance?

A. The Shift Manager may waive the pre-requisite and justify the reason in the Unit Log.

B. The Field Support Supervisor may waive the pre-requisite and must make an entry in the controlling surveillance document.

C. The I&C Supervisor may waive the pre-requisite and write a CARD to document and investigate the waiver.

D. The Shift Technical Advisor may waive the pre-requisite and document the waiver on the Surveillance Performance Form (SPF).

Answer: A ILT 2020 Final Version Page: 173 of 259 Question 68 Approved View

Answer Explanation:

MGA03, step 5.1.7, states the Shift Manager/Control Room Supervisor can waive the prerequisite if warranted. The reason for waiving the prerequisite shall be justified and documented either in the controlling document or Unit Log.

Distractor Explanation:

B. Is incorrect because the FSS cannot waive a prereq, but is plausible because the FSS role is filled by licensed SROs at Fermi 2 and the waiver could be documented on the controlling surveillance document.

C. Is incorrect because a CARD is not used to document the waiver per MGA03, but is plausible because the I&C Supervisor is the technician's supervisor.

D. Is incorrect because the waiver cannot be documented on the SPF per MGA03, but is plausible because the STAs at Fermi 2 are licensed SROs.

Reference Information:

MAG03 (5.1.7)

Plant Procedures MGA03 - Procedure Use and Adherence Question Use Closed Reference ILO RO NUREG 1123 KA Catalog Rev. 2 G2.2.12 Knowledge of surveillance procedures 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental RO Associated objective(s):

Admin Procedures Exercise Cognitive Terminal Given a condition or scenario, Describe the process for managing human performance., in accordance with the approved Fermi 2 Conduct Manuals.

ILT 2020 Final Version Page: 174 of 259 Question 68 Approved View

69 K/A Importance: 4.0/4.7 Points: 1.00 R69 Difficulty: 2.00 Level of Knowledge: Low Source: NEW 90167 Which of the following is addressed in BOTH Technical Specifications Section 2.0 Safety Limits AND Section 3.0 Limiting Conditions for Operation?

A. RCS Operational Leakage.

B. Reactor Steam Dome Pressure.

C. Linear Heat Generation Rate (LHGR).

D. Spent Fuel Storage Pool Water Level.

Answer: B ILT 2020 Final Version Page: 175 of 259 Question 69 Approved View

Answer Explanation:

Of the items listed, only Reactor Steam Dome Pressure is in Technical Specifications section 2.0, Safety Limits (2.1.2 Reactor Coolant System Pressure SL) and Section 3.0, LCOs (3.4.11 Reactor Steam Dome Pressure).

Distractor Explanation:

Distractors are incorrect and plausible because:

A. RCS Leakage is plausible because it is an LCO, 3.4.4, and it relates to the Safety Limit for Reactor Coolant System Pressure (2.1.2) in that they are both concerned with maintaining the integrity of the RCS. This distractor is incorrect because RCS Leakage is not a Safety Limit.

C. LHGR is plausible because it is an LCO, 3.2.3, and it relates to the Safety Limit for Minimum Critical Power Ratio (2.1.1.2) in that they are both concerned with maintaining the integrity of the Fuel Clad Barrier. This distractor is incorrect because LHGR is not a Safety Limit.

D. Spent Fuel Storage Pool Water Level is plausible because it closely relates to LCO 3.9.6 for RPV Water Level and SL 2.1.1.3 for Reactor vessel water level. This distractor is incorrect because Spent Fuel Storage Pool Water level is only an LCO and not a Safety Limit.

Reference Information:

Technical Specifications NUREG 1123 KA Catalog Rev. 2 G2.2.22 Knowledge of limiting conditions for operations and safety limits 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

Fermi 2 Technical Specifications Performance Enabler Perform TS Required Actions for Multiple Condition entry ILT 2020 Final Version Page: 176 of 259 Question 69 Approved View

70 K/A Importance: 3.5 Points: 1.00 R70 V2 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 91948 Consider a typical ITE 480V and ITE 4160V Breaker when answering the following question.

Assuming the Control Power Circuit is energized, and the Closing Spring Motor Disconnect Switch is on for each type of breaker, what is the status of the Closing Springs for a CLOSED 480V or 4160V ITE circuit breaker?

The closing springs are charged for __(1)__ 480V Breakers when the breakers are CLOSED.

The closing springs are charged for __(2)__ 4160V Breakers when the breakers are CLOSED.

A. (1) Some (2) Some B. (1) No (2) All C. (1) All (2) Some D. (1) Some (2) All Answer: D ILT 2020 Final Version Page: 177 of 259 Question 70 Approved View

Answer Explanation:

ILT 2020 Final Version Page: 178 of 259 Question 70 Approved View

NOTE: Drawings for both typical 4160V and 480V breakers are provided during this question to require the candidate to obtain the correct drawing to answer the 4160V portion and then obtain the correct drawing to answer the 480V portion of the question. If the candidate only refers to one drawing, and then assumes for example that this applies to both types of breakers, the candidate would answer the question incorrectly. This meets the K/A by requiring the candidate to obtain and interpret both drawings correctly.

The candidate will have to refer to (obtain) the correct schematic, for a typical 4160V and 480V breaker, to correctly determine the status of the Closing Springs for a closed breaker of the applicable type.

NOTE: These are generic explanations that are applicable to all 480V and 4160V ITE type breakers at Fermi 2. They are not specific to any one system. Both schematics will be provided during the exam.

Obtaining and referring to the wrong schematic will result in the candidate incorrectly answering the question, since the 480V and 4160V breaker closing springs do not respond in the same manner when the respective breaker is closed.

For a typical 480V Breaker (refer to SD-2548-03): When control power is available, the breaker charging motor is energized, which charges the closing springs. When the closing springs are charged, limit switch contacts 'LS1' and 'LS3' open and limit switch contact 'LS2' closes. Completion of the breaker closing circuit either by operation of the manual close control switch or by permissive logic contacts for automatic operation energizes the latch release coil, 'X' through the circuit breaker auxiliary switch 'b' contact. The latch release coil, '52X' releases the closing latch. The springs then discharge to close the circuit breaker. When the springs discharge, limit switch contacts 'LS1' and 'LS3' close and limit switch contact 'LS2' opens. Closing of the limit switch 'LS1' arms the spring charging motor circuit to recharge the closing springs as soon as the breaker has been tripped open. However, some breakers (listed in Table 2) have been modified by NOTE 10, which states "Closing springs are normally charged when the circuit breaker is open, with the exception of circuit breakers listed in Table 2. For these breakers, closing springs are charged after breaker is closed by ...."

From this, the candidate must determine that, for a typical closed 480V Breaker, (1) Some charging springs are charged.

For a typical 4160V Breaker ( refer to SD-2548-04): When control power is available, the spring charging motor is energized, which in turn charges the closing springs. When the closing springs are charged, limit switch contacts LSb are opened and limit switch contact LSa is closed. Completion of the breaker closing circuit either by operation of the manual close control switch or by permissive logic contacts for automatic closing energizes the latch release coil (x) through the circuit breaker auxiliary switch (b) contact, the normally closed lockout relay contact 'Yb' and the limit switch contact LSa. The latch release coil (x) releases the closing latch. The springs then discharge to close the circuit breaker. When the springs discharge, limit switch contacts LSb close and limit switch contact LSa opens. Closing of the limit switch contact LSb in the motor circuit energizes the spring charging motor which in turn re-charges the closing springs.

From this, the candidate must determine that, for a typical closed 4160V Breaker, (2) All charging springs are charged.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This combination of responses is possible if the candidate struggles with interpreting the schematics and defaults to deciding that some charging springs would get charged when the respective circuit breakers are closed. This is plausible because some 480V breakers do have their charging springs charged upon breaker closing. This also makes sense because having the closing springs charged when the breaker is closed means the potential energy is available, for one more closing sequence, if the breaker were to trip open and control power was not available to re-close the breaker (the breaker could be re-closed locally by depressing the local closed PB without having to charge the closing springs). This would be desirable for ESF loads but maybe not so for BOP loads (which is why the 480V breakers are different. i.e., all breakers that are different are powered by ESF busses). This is incorrect because typical 4160V breakers all have charged closing springs when the breakers are closed, and the prints show no exceptions Fermi 2.

ILT 2020 Final Version Page: 179 of 259 Question 70 Approved View

B. (1) This part is incorrect because some 480V charging springs get charged, at Fermi 2, when the respective 480V breakers are closed. This distractor is plausible because it was true prior to 2017 when the control circuit for some 480V breakers (listed in Table 2 of SD-2548-03) were modified and the candidate could fail to recognize the significance of Note 10 on the schematic. (2) This part is correct as All 4160V charging springs are charged when the respective breakers are closed at Fermi 2.

C. This combination of responses is possible if the candidate refers to the incorrect drawing for the 480V and 4160V breakers, thus reversing how each will respond. This distractor is incorrect because the responses are reversed.

Reference Information:

SD-2548-03, INTERNAL SCHEMATIC DIAGRAM ITE CIRCUIT BREAKER K-LINE TYPE 480V SWITCHGEAR SD-2548-04, INTERNAL SCHEMATIC DIAGRAM I-T-E CIRCUIT BREAKER TYPE "5HK-350&-250" 4160 V. SWITCHGEAR NUREG 1123 KA Catalog Rev. 2 G2.2.41 3.5/3.9 Ability to obtain and interpret station electrical and mechanical drawings 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Fundamental New NRC Early Review Reference Provided RO Associated objective(s):

ILT 2020 Final Version Page: 180 of 259 Question 70 Approved View

71 K/A Importance: 3.2 Points: 1.00 R71 Difficulty: 3.00 Level of Knowledge: High Source: BANK 91947 An accident has occurred which requires life saving measures.

Radiation levels in the area of the injured person are 15,000 mrem/hr.

Emergency exposure TEDE (Whole Body) limit for life saving operations has been authorized.

The MAXIMUM stay time for a rescuer under these circumstances is _____________.

A. 20 minutes B. 40 minutes C. 60 minutes D. 100 minutes Answer: D ILT 2020 Final Version Page: 181 of 259 Question 71 Approved View

Answer Explanation:

The examinee will have to recall the allowable Emergency Dose Limit to facilitate performance of the calculation below.

Per EP-201-03 Table 3, the dose limit to save lives or protect large populations is 25 REM TEDE Whole Body. 25000/15000 = 1.66 (60 mins) = 100 mins.

Distractor Explanation:

A. 5000/15000 (60 mins) = 20 mins. This distractor is plausible if the examinee determined that the dose limit for the conditions described in the stem of the question was the same as the Federal Occupational Dose Limit (Annual) given in EP-201-03 Table 1.

B. 10000/15000 (60 mins) = 40 mins. This distractor is plausible if the examinee determined that the dose limit for the conditions described in the stem of the question was 10 REM TEDE, which is plausible because 10 REM TEDE is the whole body dose limit to mitigate an accident or protect valuable property per EP-201-03 Table 2.

C. 15000/15000 (60 mins) = 60 mins. This distractor is plausible if the examinee determined that the dose limit for the conditions described in the stem of the question was 15 REM TEDE, which is plausible because 15 REM TEDE is the Federal Occupational Dose Limit (Annual) for the Lens of the Eye given in EP-201-03, Table 1.

Reference Information:

EP-201-03, Variances from Routine Radiological Practice and Procedures During an Emergency, Table 3 (pg 6)

Plant Procedures EP-201-03 NUREG 1123 KA Catalog Rev. 2 G2.3.4 Knowledge of radiation exposure limits under normal and emergency conditions 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(12) Radiological safety principles and procedures.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level RO Associated objective(s):

ILT 2020 Final Version Page: 182 of 259 Question 71 Approved View

72 K/A Importance: 3.2 Points: 1.00 R72 Difficulty: 2.00 Level of Knowledge: Fund Source: NEW 92867 You are a licensed operator performing a surveillance at a panel in the Main Control Room.

You are concerned that your Self-Reading Dosimeter (SRD) may inadvertently contact switches on the panel at which you are currently performing the surveillance.

Which of the following is allowed in accordance with MRP04, Accessing and Working in the Radiologically Controlled Area (RCA)?

A. You must contact RP to obtain permission to reposition your dosimetry.

B. You may remove your SRD in the Main Control Room since it is a designated clean area approved by the RPM.

C. You may move your dosimetry to your belt but must ensure it is repositioned correctly prior to exiting the Main Control Room.

D. You may place your SRD on a desk near you, but your SRD must be retrieved and properly repositioned prior to exiting the Main Control Room.

Answer: C ILT 2020 Final Version Page: 183 of 259 Question 72 Approved View

Answer Explanation:

Note: This question was written using lessons learned from an event in 2019 whereby an operator incorrectly placed his/her SRD on a desk and then left the MCR, at the end of shift, without the SRD.

See Corrective Action document 19-25711. This is a Rad Safety principle that applies ONLY to licensed operators because it involves a specific requirement in MRP04 for personnel working in the Main Control Room.

Per MRP04, Personnel working in the Main Control Room Complex may move dosimetry to their belt to prevent inadvertent contact with panels or equipment switches/buttons. Personnel are required to reposition dosimetry in accordance with this enclosure when exiting the Main Control Room Complex.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Permission to reposition dosimetry must be obtained from RP for personnel working on energized equipment, so the candidate could incorrectly recall that RP permission is also required to reposition dosimetry in the MCR. This is incorrect because MRP04 allows personnel in the MCR to move their dosimetry to their belt.

B. The MCR is a designated clean area inside the RCA. Because of this designation, certain activities (such as eating and drinking), that are normally prohibited in the RCA are allowed inside the MCR.

This could lead the candidate to believe that they are authorized to remove their SRD to conduct the surveillance since they are inside this designated clean area. This is incorrect because, although the MCR is designated as a clean area, the designation does not change the requirements for placement of dosimetry.

D. This is the incorrect action taken by the operator in 2019. This is incorrect because it is in violation of the requirements of MRP04.

Reference Information:

MRP04, Accessing and Working in the Radiologically Controlled Area (RCA), Enclosure F, Guidelines for Dosimetry Placement.

CARD 19-25711.

NUREG 1123 KA Catalog Rev. 2 G2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(12) Radiological safety principles and procedures.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

ILT 2020 Final Version Page: 184 of 259 Question 72 Approved View

73 K/A Importance: 3.2 Points: 1.00 R73 Difficulty: 2.00 Level of Knowledge: Fund Source: NEW 92608 You are the Control Room LNO. The plant has entered the Emergency Operating Procedures (EOPs).

You are performing actions at the H11-P601 panel in the front of the Main Control Room when an alarm is received on H11-P808 in the back of the Main Control Room.

When you scan the panels, you notice an annunciator window, with a small red frame placed around it, is flashing on H11-P808.

You cannot leave your location because of the evolution you are performing.

How should you proceed regarding this new alarm?

Silence the alarm and ...

A. Inform the CRS of the EOP entry condition.

B. Inform the CRS of a possible EOP entry condition.

C. Wait until you can get to the H11-P808 panel to evaluate.

D. Assign someone to validate the reading for entry into the EOP.

Answer: B ILT 2020 Final Version Page: 185 of 259 Question 73 Approved View

Answer Explanation:

Per ODE-10, Section on Entry Condition Identification:

Annunciators that indicate an EOP Entry Condition will be marked with a small red frame placed around the Annunciator window. Some annunciator setpoints are more conservative than the EOP Entry Condition and will also be marked with a red frame. This will indicate that an EOP Entry Condition is being approached.

The response to these alarms should be to validate the reading, using multiple indications when necessary and announce the alarm and identify if it is or not an EOP entry. If the reading cannot be verified at the time the annunciator is received then the announcement to these alarms will be modified with the phrase possible EOP entry condition. The CRS should then prioritize and assign someone to validate the reading for entry into the EOP. For example, 8D37 would be announced as Torus Water Temperature Trouble, possible EOP entry condition."

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This is the preferred way to announce red-bordered alarms. However, per ODE-10, if the reading cannot be verified at the time the annunciator is received then the announcement will be modified with the phrase "possible EOP entry condition."

C. ODE-10 says the response to red-bordered alarms should be to validate the reading, using multiple indications when necessary, and announce the alarm. However, since the stem of the question says the LNO cannot leave the front panels, ODE-10 states that the CRS should still be notified at the time the alarm is received and then it becomes the CRS's responsibility to assign someone to validate the reading.

D. The CRLNO could delegate another LNO to verify the status of the alarm, and the CRLNO may in fact ask another operator to go check the parameter as a good teamwork practice. However, for red-bordered alarms, ODE-10 assigns this responsibility to the CRS to "prioritize and assign someone to validate the reading for entry into the EOP" and not the CRLNO (or other LNO observing a red-bordered alarm).

Reference Information:

ODE-10, Emergency Operating Procedure Expectations.

NUREG 1123 KA Catalog Rev. 2 G2.4.13 4.0/4.6 Knowledge of crew roles and responsibilities during EOP usage 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

Introduction to Emergency Operating Procedures Cognitive Terminal Given a set of plant parameters that meet the entry conditions, Explain the functions performed by the Shift manager, Control Room Supervisor, Nuclear Supervising Operators, Nuclear Operators, and Shift Technical Advisor during implementation of the Emergency Operating Procedures, in accordance with Fermi 2 Emergency Operating Procedures ILT 2020 Final Version Page: 186 of 259 Question 73 Approved View

ILT 2020 Final Version Page: 187 of 259 Question 73 Approved View

74 K/A Importance: 3.1 Points: 1.00 R74 Difficulty: 2.00 Level of Knowledge: Fund Source: NEW 92443 Per MOP10, Fire Brigade:

(1) A Fire Brigade of at least how many members shall be maintained on site at all times?

(2) Which of the following, assuming he/she is qualified as a Fire Brigade Member, could be used to fill a vacancy on the Fire Brigade?

A. (1) Four (2) Security Officer B. (1) Four (2) Fire Protection Inspector C. (1) Five (2) Security Officer D. (1) Five (2) Fire Protection Inspector Answer: D ILT 2020 Final Version Page: 188 of 259 Question 74 Approved View

Answer Explanation:

Per MOP10, Fire Brigade, A Fire Brigade of at least five members shall be maintained onsite at all times.

The Fire Brigade shall not include the SM, Security, or members of the minimum shift crew necessary for safe shutdown of the unit or any personnel required for other essential functions during a fire emergency.

Paragraph 3.1.1 states that a typical Fire Brigade should consist of:

1. One Fire Brigade Leader.

2. Four Operators (or three Operators and a Fire Brigade Qualified Fire Protection Inspector)

3. One other qualified member of plant staff as communicator.

Distractor Explanation:

Distractors are incorrect and plausible because:

(1) Four is plausible if the examinee failed to consider the Fire Brigade Leader as a member of the Fire Brigade. However, MOP10 requires that at least five members shall be maintained onsite at all times.

(2) Security Officer is plausible because they perform rounds throughout the plant, have access to all plant areas, and often fill roles to walkdown areas with inoperable fire detection/protection equipment.

Also, it was previously allowed for a Security Officer to serve on the Fire Brigade, so an examinee who is studying an out of date exam or related study material, could plausibly choose this option.

However, a commitment has been made (CM 20079) to not use Security Officers on the Fire Brigade, as spelled out in MOP10.

Reference Information:

MOP10, Fire Brigade NUREG 1123 KA Catalog Rev. 2 G2.4.26 Knowledge of facility protection requirements including fire brigade and portable fire fighting equipment usage 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New RO Associated objective(s):

Admin Procedures Exercise Cognitive Terminal Given a condition or scenario, Describe the typical Fire Brigade composition and any exceptions to the minimum requirements, in accordance with the approved Fermi 2 Conduct Manuals.

ILT 2020 Final Version Page: 189 of 259 Question 74 Approved View

75 K/A Importance: 3.6/4.0 Points: 1.00 R75 Difficulty: 3.00 Level of Knowledge: High Source: BANK SOURCE: 2018 92218 NRC EXAM The plant is currently operating at 100% power in a black board condition when the CRLNO notices that the VAS display shows 3D19 as red, along with VAS Hardware System Trouble and MUX A & C FAILED.

Which of the responses below correctly completes the following statement regarding the impact of these conditions on the VAS system?

This information indicates that the VAS system will incur ____(1)____of functionality and will result in _____(2)_____.

A. (1) loss (2) loss of redundancy for one half side of VAS I/O B. (1) no loss (2) loss of P601 through P805 window failures C. (1) loss (2) loss of P601 through P805 window failures D. (1) no loss (2) loss of redundancy for one half side of VAS I/O Answer: C ILT 2020 Final Version Page: 190 of 259 Question 75 Approved View

Answer Explanation:

NOTE: ARP 3D19 to be provided with this question.

ARP 3D19 includes a flowchart on Page 6 that the candidate should refer to based on the information provided in the stem of the question.

Due to the VAS System Hardware Failed and the blackboard condition, the operator should answer N to the question "have a number of false alarm windows turned on?" This should lead the candidate to review the table to determine Data Acquisition Status. Given the failure of MUX A and C, the operator should review Action C, which is located on the next page (page 7). This page describes the impact of a loss of the redundant pair of MUX A&C.

A failure of this pair (or any redundant pair, which includes A&C or B&D) will result in a loss of VAS functionality and loss of alarm windows on the P601 through P805 panels.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The candidate could incorrectly read the flowchart in the ARP and read Action B, which would lead the candidate to determine that the loss MUX A/C means that redundancy for one half side of VAS was lost, which is incorrect because the redundant pair has been lost (A and C MUX are redundant to one another).

B. The candidate could incorrectly read the flowchart in the ARP and determine that no loss of functionality occurred, which is incorrect as stated on Action C.

D. The candidate could incorrectly read the flowchart in the ARP and determine that no loss of functionality occurred, which is incorrect as stated on Action C.

Reference Information:

ARP 3D19, Annunciator System Trouble (provided as reference)

NUREG 1123 KA Catalog Rev. 2 G2.4.32 Knowledge of operator response to loss of all annunciators 10CFR55 RO/SRO Written Exam Content 10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fermi 2 NRC Exam Usage ILO 2018 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Higher Cognitive Level Reference Provided RO Associated objective(s):

Visual Annunciator System (VAS)

Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: Identify alarm response procedures associated with the Visual Annunciator System.

ILT 2020 Final Version Page: 191 of 259 Question 75 Approved View

76 K/A Importance: 4.4 Points: 1.00 S76 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 87251 You are the Control Room Supervisor. The plant is operating at 100% power with the following conditions:

N RRMG Set Speed: 74.6%.

S RRMG Set Speed: 74.6%.

B21-R611A, Jet Pump 11-20 Loop A Flow: 44.7 x 106 LB/Hr.

B21-R611B, Jet Pump 1-10 Loop B Flow: 44.7 x 106 LB/Hr.

B21-R613, Reactor Jet Pump Total Flow: 89.43 x 106 LB/Hr.

Several alarms are subsequently received on the H11-P603 panel and the following are observed:

N RRMG Set Speed: 64% and lowering.

S RRMG Set Speed: 74.6% and steady.

B21-R611A, Jet Pump 11-20 Loop A Flow: 39.0 x 106 LB/Hr and lowering rapidly.

B21-R611B, Jet Pump 1-10 Loop B Flow: 45.0 x 106 LB/Hr and rising slowly.

B21-R613, Reactor Jet Pump Total Flow: 83.0 x 106 LB/Hr and lowering rapidly.

(1) Which of the following actions will you ensure is carried out by the P603 Operator?

(2) What action(s), if any, is(are) required by Technical Specifications?

A. (1) Lock the Scoop Tube for the N RRMG Set.

(2) No Tech Spec required actions are necessary.

B. (1) Trip the N RRMG Set.

(2) Apply limitations for single loop operation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C. (1) Lock the Scoop Tube for the N RRMG Set.

(2) Declare Recirculation Loop A not in operation within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

D. (1) Trip the N RRMG Set.

(2) Declare Recirculation Loop A not in operation within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months then apply limitations for single loop operation within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

Answer: C ILT 2020 Final Version Page: 192 of 259 Question 76 Approved View

Answer Explanation:

For part (1), per 20.138.01, Immediate Actions, the correct action to take per IA is to Lock the scoop tube for the affected RRMG Set, which is RRMG Set A.

For part (2), per Technical Specifications for LCO 3.4.1, SR 3.4.1.1, loop jet pump flows must be within 5% when operating at 70% core flow. The candidate must evaluate core flows and determine that (1) total core flow is 83% and (2) jet pump flow mismatch is 6%, thus making SR 3.4.1.1 NOT MET.

Therefore, the candidate must recall that LCO 3.4.1 Condition A requires declaring the recirculation loop with lower flow (in this case, Loop A) not in operation within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

Distractor Explanation:

Distractors are incorrect and plausible because:

A. (1) This part is correct. (2) This part is plausible because of the confusion (common misconception made) regarding the requirements for matched loop flows in SR 3.4.1.1. This SR requires the two loop jet pump flows to be within 5% when operating with rated core flow 70%. It is common for licensed operators to apply the larger allowable mismatch (10%) to the higher rated core flow (70%).

This distractor is incorrect because a lower mismatch (5%) is allowed with higher rated core flow (70%), therefore the SR is not met.

B. (1) This part is plausible because Immediate Action IB requires tripping the affected RRMG set when its speed has increased 10%. Since the conditions in the stem indicate a pump decrease 10%, the candidate could incorrectly determine that the correct action is to trip the RRMG set with the changing speed. This is incorrect because IB only requires tripping the RRMG set if speed has increased 10%. (2) This part is plausible because this is the correct action, required by Technical Specifications, for a tripped recirculation pump. However, since the AOP does not require tripping the RRMG set, this distractor is incorrect.

D. (1) This part is plausible because Immediate Action IB requires tripping the affected RRMG set when its speed has increased 10%. Since the conditions in the stem indicate a pump decrease 10%, the candidate could incorrectly determine that the correct action is to trip the RRMG set with the changing speed. This is incorrect because IB only requires tripping the RRMG set if speed has increased 10%. (2) This part is plausible because Condition A allows 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months before declaring a recirculation loop with lower flow not in operation so the candidate might add these 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months allowed by the note in the LCO statement before taking action for transitioning to single loop operations. This is incorrect because Condition A only applies to jet pump loop flows not within mismatch limits and is not applicable for a tripped recirculation loop.

10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires application of required actions (>1 hour) and knowledge of surveillance requirements which are not listed "above the line".

The question cannot be answered solely by knowing 1-hour TS/TRM Actions, or solely by knowing information in TS/TRM listed above the line or solely by knowing TS Safety Limits.

Reference Information:

20.138.01, Uncontrolled Recirc Flow Change AOP.

Technical Specifications LCO 3.4.1, Recirculation Loops Operating ILT 2020 Final Version Page: 193 of 259 Question 76 Approved View

NUREG 1123 KA Catalog Rev. 2 295001 Partial or Complete Loss of Forced Core Flow Circulation G2.4.49 Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New SRO Associated objective(s):

Reactor Recirculation System Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: Describe the Reactor Recirculation system technical specification limiting conditions for operation, their bases, the associated surveillance requirement(s), and their relationship to operability.

ILT 2020 Final Version Page: 194 of 259 Question 76 Approved View

77 K/A Importance: 3.8 Points: 1.00 S77- Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 97307 POST SUBMIT TAL VERSIO N

Consider the design basis transient analysis for the feedwater controller failure to maximum demand event to answer this question.

If this event were to occur above 30% power:

(1) Which of the following is assumed in the analysis of the event to cause a reactor scram?

(2) The reactor scram mitigates the reduction in what thermal limit?

A. (1) Trip of the Main Turbine.

(2) Linear Heat Generation Rate (LHGR).

B. (1) Trip of the Main Turbine.

(2) Minimum Critical Power Ratio (MCPR).

C. (1) High Reactor Pressure.

(2) Linear Heat Generation Rate (LHGR).

D. (1) High Reactor Pressure.

(2) Minimum Critical Power Ratio (MCPR).

Answer: B ILT 2020 Final Version Page: 195 of 259 Question 77 Approved View

Answer Explanation:

Per TS Bases for LCO 3.3.2.2, Feedwater and Main Turbine High Water Level Trip Instrumentation, Applicable Safety Analyses: The feedwater and main turbine high water level trip instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event. The Level 8 trip indirectly initiates a reactor scram (above 29.5% RTP) from the main turbine trip and trips the feedwater pumps, thereby terminating the event. The reactor scram mitigates the reduction in MCPR.

Therefore, the SRO examinee must recall that it is the trip of the Main Turbine that causes a scram to terminate the event and that the scram terminates the reduction in the MCPR thermal limit.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. (1) is correct. (2) is plausible because LHGR is one of the three thermal limits applicable to Fermi 2 and LHGR, like MCPR, is a transient limit (unlike APLHGR, which is an accident limit) that is assumed in numerous accident analyses for which several instruments are taken credit. This answer is incorrect because the accident analysis for the feedwater controller to maximum demand event assumes that the Main Turbine trip mitigates the reduction in MCPR not LHGR.

C. (1) is plausible because High Reactor Pressure is an RPS Scram signal and high reactor pressure will result from trip of the Main Turbine when the turbine trips on High RPV Water Level (Level 8). The candidate could incorrectly recall that it is High Reactor Pressure that is assumed in the safety analysis since the safety analysis states that it relies on an indirect scram to terminate the event (no direct scram occurs from reaching L8). (2) is plausible because LHGR is one of the three thermal limits applicable to Fermi 2 and LHGR, like MCPR, is a transient limit (unlike APLHGR, which is an accident limit) that is assumed in numerous accident analyses for which several instruments are taken credit. This answer is incorrect because the accident analysis for the feedwater controller to maximum demand event assumes that the scram is caused by the Main Turbine trip, which is assumed to mitigate the reduction in MCPR.

D. (1) is plausible because High Reactor Pressure is an RPS Scram signal and high reactor pressure will result from trip of the Main Turbine when the turbine trips on High RPV Water Level (Level 8). The candidate could incorrectly recall that it is High Reactor Pressure that is assumed in the safety analysis since the safety analysis states that it relies on an indirect scram to terminate the event (no direct scram occurs from reaching L8). Part (2) is correct. This answer is incorrect because the accident analysis for the feedwater controller to maximum demand event assumes that the scram is caused by the Main Turbine trip, not low RPV level due to the RFP trip.

10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires knowledge of TS bases that is required to analyze TS terminology.

The question cannot be answered solely by knowing 1-hour TS actions, or information above the line or TS safety limits.

Reference Information:

TS Bases for LCO 3.3.2.2, Feedwater and Main Turbine High Water Level Trip Instrumentation.

ILT 2020 Final Version Page: 196 of 259 Question 77 Approved View

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

295006 AA2.06 3.5/3.8 Cause of reactor SCRAM 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New SRO Associated objective(s):

ILT 2020 Final Version Page: 197 of 259 Question 77 Approved View

78 K/A Importance: 4.3 Points: 1.00 S78 Difficulty: 3.00 Level of Knowledge: High Source: NEW 91847 You are the Control Room Supervisor (CRS). The plant is in MODE 4 with heat input to the reactor coolant greater than heat losses to ambient.

All plant equipment is available.

Division 1 RHR is running in the Shutdown Cooling Mode.

RHR Pump C is running.

The E1150-F010, RHR Cross-Tie Valve, is open.

While walking down the control room panels, you notice all lights out on the CMC switch for the A RHR Pump. The CRLNO replaced all light bulbs, but the lights remained out.

While investigating the above, the C RHR Pump trips for an unknown reason.

(1) How many RHR Shutdown Cooling subsystems are currently OPERABLE?

(2) What Action should you direct to restore RHR Shutdown Cooling to operation?

A. (1) One.

(2) Start RHR Pump B.

B. (1) Two.

(2) Start RHR Pump B.

C. (1) One.

(2) Remove Division 1 RHR from service and place Division 2 RHR in SDC.

D. (1) Two.

(2) Remove Division 1 RHR from service and place Division 2 RHR in SDC.

Answer: D ILT 2020 Final Version Page: 198 of 259 Question 78 Approved View

Answer Explanation:

The SRO examinee must evaluate the control room indications for RHR Pump A and determine that a loss of control power must have occurred, which makes RHR Pump A unavailable for use in the SDC mode.

Per Tech Spec BASES B 3.4.9, Each RHR shutdown cooling loop consists of two subsystems. An OPERABLE RHR shutdown cooling subsystem consists of one OPERABLE RHR pump, one heat exchanger, and the associated piping and valves. The two subsystems have a common suction source and are allowed to have a common heat exchanger and common discharge piping.

Therefore, the SRO examinee must recall this definition from Tech Spec Bases and determine that two (2) RHR Shutdown Cooling sub-systems are OPERABLE because Division 2 RHR is available.

Given that no sub-system in the Division 1 RHR SDC Loop is OPERABLE, the correct course of action is to direct Actions H.1 and H.2 to Remove Division 1 RHR from service and Place Division 2 RHR in SDC to operation to meet LCO 3.4.9.

Note: When considering the plausibility of the distractors below, consider the TS BASES definition of a LPCI subsystem from B 3.5.1, which states There are two LPCI sub-systems, each consisting of two motor driven pumps and piping and valves to transfer water from the suppression pool to the RPV. It is a common misconception among SRO candidates to assume that a sub-system in LCO 3.4.9 requires 2 RHR pumps by confusing the BASES for LCO 3.5.1 with that of LCO 3.4.9.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The SRO examinee could incorrectly recall that, to meet the LCO, both pumps in each of the two loops must be OPERABLE and, therefore, determine that, with Division 1 RHR unavailable, only ONE sub-system (Division 2) is OPERABLE. This is not correct, since the two Division 2 pumps (and piping, HX, etc.) satisfy the LCO requirement for 2 sub-systems, not one. The examinee could then conclude that, with the E1150-F010 open, all that is necessary to restore SDC to operation is to start a Division 2 RHR pump. This is incorrect because Condition H of the AOP is applicable (RHR SDC Flow cannot be restored in loop that was in service) so the correct action is to remove Div 1 and place Div 2 RHR in SDC.

B. The examinee could correctly determine the number of OPERABLE SDC sub-systems but incorrectly conclude that, with the E1150-F010 open, all that is necessary to restore SDC to operation is to start a Division 2 RHR pump. This is incorrect because Condition H of the AOP is applicable (RHR SDC Flow cannot be restored in loop that was in service) so the correct action is to remove Div 1 and place Div 2 RHR in SDC.

C. The SRO examinee could incorrectly recall that, to meet the LCO, both pumps in each of the two loops must be OPERABLE and, therefore, determine that, with Division 1 RHR unavailable, only ONE sub-system (Division 2) is OPERABLE. This is not correct, since the two Division 2 pumps (and piping, HX, etc.) satisfy the LCO requirement for 2 sub-systems, not one. The second part is correct.

10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires knowledge of TS bases that are required to analyze TS-required actions and terminology.

The question cannot be answered solely by knowing </= 1-hour TS/TRM Actions, or solely by knowing LCO/TRM information listed above the line, or solely by knowing the TS safety limits.

Reference Information:

Tech Spec BASES B 3.4.9, RHR Shutdown Cooling System-Cold Shutdown.

20.205.01, Loss of Shutdown Cooling AOP.

ILT 2020 Final Version Page: 199 of 259 Question 78 Approved View

NUREG 1123 KA Catalog Rev. 2 295021 Loss of Shutdown Cooling G2.1.31 Ability to locate control room switches, controls and indications and to determine that they are correctly reflecting the desired plant lineup 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New SRO Associated objective(s):

Residual Heat Removal Cognitive Terminal Given the system operating conditions/parameters, in accordance with approved plant procedures: Describe the Residual Heat Removal System technical specification limiting conditions for operation, their bases, the associated surveillance requirement(s), and their relationship to operability.

ILT 2020 Final Version Page: 200 of 259 Question 78 Approved View

79 K/A Importance: 4.1 Points: 1.00 S79- Difficulty: 4.00 Level of Knowledge: High Source: MODIFIED 97327 POST SUBMIT TAL VERSIO N

You are the Control Room Supervisor.

The EOPs were entered due to a Torus Leak. The leak was stopped by isolating Division 2 RHR.

A subsequent LOCA in the Drywell has caused a scram on High Drywell Pressure.

RPV Level is being maintained by Core Spray.

Containment parameters are:

T23-R800, Torus Water Temperature Recorder, Point 9 ...... 190°F.

T50-R800A/B, Div 1/2 PC Air and Water Temp Rec, Points 11/12 ..... 180°F.

T47-R803A/B, Div 1/2 Drywell Temperature Rec, Point 24 ... 245°F.

T50-R802A/B, Div 1/2 Drywell Pressure .... 12 psig.

T50-R802A/B, Div 1/2 Torus Pressure ,,,.. 2 psig.

T50-R804A/B, Div 1/2 Suppression Chamber Water Level ...... -25.

Which of the following will you direct?

A. Place RHR Pump A or C in Torus Cooling and Torus Spray ONLY, with flow limited to 7,000 gpm.

B. Place RHR Pump A or C in Torus Cooling and Torus Spray ONLY, with flow limited to 10,000 gpm.

C. Place RHR Pumps A and C in Torus Cooling, Torus Spray AND Drywell Spray, with flow limited to 14,000 gpm.

D. Place RHR Pumps A and C in Torus Cooling, Torus Spray AND Drywell Spray, with flow limited to 20,000 gpm.

Answer: D ILT 2020 Final Version Page: 201 of 259 Question 79 Approved View

Answer Explanation:

Note: This question was modified from Q77 of the Limerick 2017 NRC Exam.

Note: 29.100.01, Sheet 6 will be provided with this question.

This question matches the K/A because the candidate must first recall the Drywell Pressure (PCP) and Drywell Temperature (DWT) legs of 29.100.01, Sheet 2, Primary Containment Control, to determine if conditions require just Torus Cooling and Torus Spray or TC, TS AND Drywell Spray. Then, the candidate must interpret the affects of elevated Torus Temperature on system flow, due to the degraded RHR configuration given in the stem, when the candidate is determining the proper RHR system configuration to combat High Drywell Pressure and Temperature.

With Drywell Pressure >1.68 psig, the candidate must determine that the PCP leg has been entered and the PCP leg requires Torus Cooling and Torus Spray be placed in service. The candidate must then determine that, with DWT >242°F, Drywell Spray is required per the DWT leg.

Once the candidate determines the proper configuration of the RHR system for the given containment conditions, the candidate must utilize Sheet 6 of the EOPs to interpret how Torus Temperature will limit RHR flow as RHR is used to combat the elevated Drywell Pressure and Temperatures. Per 29.100.01, Sheet 6, the examinee will have to recognize that, with Torus Water Level <-11, the normal recorder for monitoring Torus Water Temperature, T23-R800, Torus Water Temperature Recorder, is unavailable.

Torus Temperature monitoring will shift to the T50-R800A/B, Div 1/2 PC Air and Water Temp Recorders per Caution 6. Therefore, the examinee must determine that the accurate Torus Water Temperature to use is 180°F.

The candidate must calculate Torus Overpressure given the information in the stem:

Torus Overpressure = (Ind Torus Press) + 3.5 psig + (Ind Torus Level / 30)

Torus Overpressure = (2 psig) + 3.5 psig + (-25/30)

Torus Overpressure = 4.7 psig.

Therefore, the candidate will have to use the 0 psig curve, because of the note prohibiting interpolation between curves and requiring the next lower curve be used.

Therefore, the SRO candidate must use the 0 psig NPSH Limit curve, at a Torus Water Temperature of 180°F, to determine that maximum flow for 2 RHRs pump (A and C) is 20,000 gpm.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. 7,000 gpm is the approximate intersection between the 0 psig curve and 190°F for 1 RHR Pump, which could be used if the examinee determined that only TC and TS is required and the candidate failed to recognize the significance of Torus Water Level being below -11. This is incorrect because TC, TS and DS are required (thus both RHR Pumps A and C) and the examinee should be using T50-R800A/B, Div 1/2 PC Air and Water Temp Rec, Points 11/12 to determine that the Torus Water Temperature to use is 180°F.

B. 10,000 gpm is the approximate intersection between the 0 psig curve and 180°F for 1 RHR Pump, which could be used if the examinee determined that only TC and TS is required. This distractor is incorrect because both RHR Pumps A and C should be placed in service, to support RHR in TC, TS and DS, therefore the 2 pump limit would apply.

C. 14,000 gpm is the approximate intersection between the 0 psig curve and 190°F for 2 RHR Pumps, which could be used if the examinee failed to recognize the significance of Torus Water Level being below -11. This distractor is incorrect because the examinee should be using T50-R800A/B, Div 1/2 PC Air and Water Temp Rec, Points 11/12 to determine that the Torus Water Temperature to use is 180°F.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires assessment of plant conditions and determining the correct EOP actions with which to proceed. Also, since Torus Level is <-11, the NPSH limit curve must be evaluated by pulling out a hard copy of the curves because the instrument the plant computer (IPCS) uses to plot position on the curves is invalid below -11, making this an SRO-Only function.

ILT 2020 Final Version Page: 202 of 259 Question 79 Approved View

The question cannot be answered solely by knowing "systems knowledge", or solely by knowing immediate operator actions, or solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs, or solely by knowing the purpose, overall sequence of events, or mitigative strategy of a procedure.

Reference Information:

29.100.01, Sheet 6, Curves, Cautions and Tables (provided).

29.100.01, Sheet 2, Primary Containment Control (not provided).

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

295024 EA2. Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE:

295024 EA2.06 4.1/4.1 Suppression pool temperature 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Higher Cognitive Level Modified Reference Provided SRO Associated objective(s):

ILT 2020 Final Version Page: 203 of 259 Question 79 Approved View

80 K/A Importance: 3.7/4.7 Points: 1.00 S80- Difficulty: 3.00 Level of Knowledge: High Source: BANK: FERMI LOR 97308 POST 2009, 2015 SUBMIT TAL VERSIO N

The plant was at 100% power when a LOCA occurred:

Reactor power ..................................................... all control rods fully inserted.

MSIVs .....................................................................................................closed.

Reactor pressure ................................................................................ 265 psig.

RPV water level ....................................................... 5 inches (slowly lowering).

Drywell temperature ................................................................... 342°F (rising).

Drywell pressure ..................................................................................... 6 psig.

Torus water level .................................................................................0 inches.

RHR Pumps A & D .............................................................. injecting into RPV.

All other RPV injection sources ....................................................... inoperable.

Which one of the actions below must the CRS direct?

A. Initiate Drywell Sprays.

B. Emergency Depressurize the RPV.

C. Restore Drywell Cooling per 29.ESP.08 D. Bypass interlocks, reopen MSIVs, and depressurize to the main condenser.

Answer: B ILT 2020 Final Version Page: 204 of 259 Question 80 Approved View

Answer Explanation:

Per 29.100.01 sheet 2, PC Control, when drywell temperature cannot be restored and maintained below 340°F, ED is required. Drywell sprays cannot be initiated since all RHR pumps are required for adequate core cooling, and no other injection sources are available.

A is incorrect because RHR should not be diverted from RPV injection. It is plausible if applicants incorrectly determine that RHR flow should be diverted to DW sprays.

C is incorrect because inside of 29.ESP.08 it states that the procedure must be stopped if DW temperature exceeds 340°F. This is plausible since DW cooling is restored, using 29.ESP.08, prior to reaching 242°F, so it is a viable strategy of primary containment protection under different conditions.

D is incorrect because ED is required, and this action does not require use of the main condenser. This is plausible because this action is appropriate under different conditions, and applicants may incorrectly determine that ED can be anticipated, but this is not allowed once ED is required.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOPs, it is not related to immediate actions, and the entry conditions are not relevant or leading to the answer. The answer to this question is based on assessing plant conditions and then applying the requirements of the EOPs.

References:

29.100.01 sheet 2, PC Control This question matches the selected K/A because SRO applicants must evaluate plant conditions and apply knowledge of EOP mitigation strategies.

29.100.01 sheet 1, RPV Control NUREG 1123 KA Catalog Rev. 2 G2.4.6 Knowledge of EOP mitigation strategies 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage LOR 2009 Exam LOR 2015 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level SRO Associated objective(s):

ILT 2020 Final Version Page: 205 of 259 Question 80 Approved View

81 K/A Importance: 4.1/4.5 Points: 1.00 S81- Difficulty: 3.00 Level of Knowledge: High Source: BANK: 2018 FERMI 2 97310 POST NRC EXAM SUBMIT TAL VERSIO N

A fuel failure has occurred, and a Radiation Release is in progress.

The Offsite rad release rate is higher than the ALERT offsite release rate and less than the GENERAL EMERGENCY release rate.

Use of HPCI is required to maintain Reactor Water Level. HPCI room temperature and radiation levels are slowly rising.

Secondary Containment has isolated due to High Radiation in the RBHVAC Exhaust Plenum.

Turbine Building HVAC radiation levels are rising and approaching the trip setpoint.

Which of the following would the CRS direct to be isolated, per the Rad Release section of the EOPs, to mitigate the High Offsite Release Rate?

A. HPCI.

B. Offgas System.

C. Main Steam Lines.

D. Turbine Building HVAC.

Answer: C ILT 2020 Final Version Page: 206 of 259 Question 81 Approved View

Answer Explanation:

Per 29.100.01 SH 5 step RR-2:

Isolate ALL radioactivity releases that are discharging into areas outside the primary and secondary containment, except systems required by the EOPs.

Each of the systems listed would be considered for isolation by the CRS since each of them could either contribute to the release of fission products into Secondary Containment or the transport of fission products outside of containment.

Isolating the Main Steam Lines is therefore correct because the fission products can exit Secondary Containment via them and they are not required for the EOPs (an ATWS is not in progress requiring the Main Condenser as a heat sink, for example).

Distractor Explanation:

Distractors are incorrect and plausible because:

A. HPCI could contribute to radioactivity release since it is a primary system with piping in Secondary Containment However, since HPCI is required by the EOPs, for maintaining RPV water level, it should not be isolated.

B. Rising offsite radiation levels may lead the candidate to conclude that isolating the Offgas System is necessary to stop further radioactivity release from the Main Condenser. However, Offgas is not a primary system and it would be preferable to maintain Offgas in service to provide a holdup for fission product gasses prior to release and to discharge any release from the Main Condenser through an elevated and monitored release point.

D. Rising radiation levels in the TBHVAC may lead the candidate to conclude that isolating TBHVAC is necessary. This is incorrect because it is preferable to maintain HVAC systems running to preserve building accessibility and to discharge radioactivity through an elevated, monitored release path (vice isolating HVAC leading to an unmonitored ground-level release). Guidance to restart HVAC systems, even those isolated due to high radiation, is given in RR-OR1 (Override 1).

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of EOPs, it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this question is based on assessing plant conditions and then applying the requirements of the EOPs.

Reference Information:

29.100.01 Sheet 5, Secondary Containment Control and Rad Release.

BWROG EPGs/SAGs Appendix B Vol. II (EPGs-Hot) Radioactivity Release Control.

ILT 2020 Final Version Page: 207 of 259 Question 81 Approved View

Plant Procedures 29.100.01 SH 5 NUREG 1123 KA Catalog Rev. 2 295038 EA2. Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE :

295038 EA2.04 Source of off-site release 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILO 2018 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level SRO Associated objective(s):

ILT 2020 Final Version Page: 208 of 259 Question 81 Approved View

82 K/A Importance: 4.1 Points: 1.00 S82 Difficulty: 3.00 Level of Knowledge: high Source: MODIFIED: COOPER 90270 2017 NRC The plant is in MODE 5 with refueling operations in progress.

Div 1 safety systems are OOS

Div 2 safety systems are operable and required Maintenance workers in the RHR complex report a fire within the EDG 14 engine room.

20 minutes later the fire brigade reports that the EDG 14 oil sump is on fire.

(1) What is the HIGHEST EAL classification?

(2) Which offsite authority is REQUIRED to be notified within 15 minutes?

A. (1) Alert (2) State of Michigan B. (1) Unusual Event (2) NRC Operations Center C. (1) Alert (2) NRC Operations Center D. (1) Unusual Event (2) State of Michigan Answer: A ILT 2020 Final Version Page: 209 of 259 Question 82 Approved View

Answer Explanation:

NOTE: Bank Source is 2017 Cooper NRC ILE (Q81).

Per EP-101 EAL charts:

In mode 5, EAL CA6.1 applies due to damage to EDG14 which is required for the current plant conditions.

CA6.1: The occurrence of any Table C-5 hazardous (fire in this case) event AND EITHER of the following:

Event damage has caused indications of degraded performance in at least one train of a SAFETY SYSTEM required for the current operating mode

The event has caused VISIBLE DAMAGE to a SAFETY SYSTEM component or structure required for the current operating mode EAL HU4.1 is also applicable in this instance, however, it's not the highest classification.

HU4.1: A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications (Note 1):

Report from the field (i.e., visual observation)

Receipt of multiple (more than 1) fire alarms or indications

Field verification of a single fire alarm AND

The FIRE is located within any Table H-1 area (RHR complex)

Per EP--290, Emergency communications, only offsite authorities in (1) Wayne Co., (2) Monroe Co., and (3) State of Michigan MUST be notified within 15 minutes.

B is incorrect since CA6.1 is the highest applicable EAL. It is plausible because HU4.1, also applies, however it is not the HIGHEST classification under current conditions.

C is incorrect since CA6.1 is the highest applicable EAL, and the NRC is not required to be notified within 15 minutes. It is plausible because HU4.1 (see description above), also applies, however it is not the HIGHEST classification under current conditions; and NUREG 1022 requires NRC notification of an EAL classification, however the time limit is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , not 15 minutes.

D is incorrect since the NRC is not required to be notified within 15 minutes. It is plausible because NUREG 1022 requires NRC notification of an EAL classification, however the time limit is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , not 15 minutes.

References:

EP-101, Classification of Emergencies EP-290, Emergency Notifications This question is SRO Only (10CFR55.43(b)(5)) because it requires assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed.

It cannot be answered solely by knowing system knowledge, immediate actions, EOP/AOP entry conditions or the overall mitigating strategy of procedures.

It matches the selected K/A since SRO applicants must assess conditions associated with a fire on site, and determine which offsite agencies must be notified within specific time periods.

ILT 2020 Final Version Page: 210 of 259 Question 82 Approved View

NUREG 1123 KA Catalog Rev. 2 600000 Plant Fire On Site G2.4.30 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 the transmission system operator 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Higher Cognitive Level Modified Reference Provided SRO Associated objective(s):

RERP: Emergency Classifications and Protective Action Recommendations Performance Enabler Classify Emergency Events ILT 2020 Final Version Page: 211 of 259 Question 82 Approved View

83 K/A Importance: 4.1/4.1 Points: 1.00 S83 Difficulty: 2.00 Level of Knowledge: Fund Source: BANK: NMP2 2010 91050 NRC EXAM An event occurred with the following conditions:

An RPV pressure transient occurred with steam dome pressure peaking at 1350 psig.

5 minutes later, RPV pressure was reduced to 920 psig, and is now stable.

Which of the following additional actions is REQUIRED to be taken by TECHNICAL SPECIFICATIONS?

A. Insert all insertable control rods within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. Initiate action to be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

C. Initiate action to be in MODE 3 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months .

D. Initiate action to insert all insertable control rods immediately.

Answer: A ILT 2020 Final Version Page: 212 of 259 Question 83 Approved View

Answer Explanation:

Per tech spec 2.1.2 the Reactor Coolant System Safety limit of 1325 psig has been exceeded, and action 2.2.2, Insert all insertable control rods within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , must be performed following restoration of pressure below the SL.

Distractor Explanation:

B is incorrect since this action is not required following a safety limit violation. Plausible since this action is required by TS 3.4.11 following exceeding 1045 psig steam dome pressure, but ONLY if pressure is not reduced <1045 within 15 minutes. In this instance, pressure has been reduced below 1045 in 5 minutes which satisfies TS 3.4.11 action A.

C is incorrect since this action is not required following a safety limit violation. Plausible since this action is required when TS 3.03 is invoked.

D is incorrect since this action is not required following a safety limit violation. Plausible since the candidate may believe that violation of a safety limit is so severe that it requires immediate control rod insertion. Also, several other TS sections require this action, such as 3.1.1 for SDM violations.

10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires application of the required TS actions following a safety limit violation. Since the actions for SL violations have a >1-hour completion time, there is no objective requiring ROs to know this information, making actions taken for SL violations SRO Only knowledge at Fermi 2.

The question cannot be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS/TRM action, LCO/TRM action above the line, or by solely knowing the safety limits.

Reference Information:

Tech Spec 2.1, 2.2, 3.4.11 NUREG 1123 KA Catalog Rev. 2 295007 High Reactor Pressure 295007 AA2. Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE :

295007 AA2.01 4.1*/4.1* Reactor pressure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental SRO Associated objective(s):

Fermi 2 Technical Specifications Performance Enabler Perform TS Required Actions for Multiple Condition entry ILT 2020 Final Version Page: 213 of 259 Question 83 Approved View

84 K/A Importance: 3.9 Points: 1.00 S84 - Difficulty: 3.00 Level of Knowledge: Source: NEW 91927 POST Fundamental SUBMIT TAL VERSIO N

Failure of (1) would require entry into LCO 3.3.3.1 PAM Instrumentation.

This PAM instrument is provided to (2) .

A. (1) B21-R610, RPV Core Level Recorder on H11-P601, Div 1 ECCS Panel.

(2) give the operator water level information covering the area of interest during an accident.

B. (1) B21-R610, RPV Core Level Recorder on H11-P601, Div 1 ECCS Panel.

(2) close the RHR test/suppression pool cooling line, suppression pool spray, and containment spray isolation valves to allow full LPCI system flow to the RPV.

C. (1) C35-R002, Div 1 Reactor Vessel Wide Range Level Indicator on H21P100, Remote Shutdown Panel.

(2) give the operator water level information covering the area of interest during an accident.

D. (1) C35-R002, Div 1 Reactor Vessel Wide Range Level Indicator on H21P100, Remote Shutdown Panel.

(2) provide the operator with instrumentation to place the plant in a safe condition from a location other than the control room.

Answer: A ILT 2020 Final Version Page: 214 of 259 Question 84 Approved View

Answer Explanation:

Per 44.030.082 LCO 3.3.3.1-1 function 1 is impacted by B21-R610, RPV Core Level Recorder on H11-P601.

Per TS 3.3.3.1 Bases:

For Functions 2 and 3, Reactor Vessel Water Level - Fuel Zone; Reactor Vessel Water Level - Wide Range:

The wide range and fuel zone range water level channels provide the PAM Reactor Vessel Water Level Function.

The two measurement systems provide overlapping ranges to give the operator water level information covering the area of interest during an accident.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. The instrument given is a PAM instrument, but the bases given is not the basis for PAM instrumentation, but for an isolation (at Level 0) that occurs from the transmitter that drives the recorder given in part (1). This is plausible because the examinee could determine that the isolation is part of the PAM function but incorrect because the PAM instrument bases is concerned with providing operator indications during accidents and not the isolations off these instruments.

C. The instrument given is an instrument located on the Remote Shutdown Panel, which is required by TS LCO 3.3.3.2. It is plausible that the candidate could determine that this is a PAM instrument because TS LCO 3.3.3.1 Table 3.3.3.1-1 requires 2 Wide Range Level Instruments to satisfy Function 3 and the instrument given in this distractor is a Wide Range Level instrument. The candidate could then incorrectly determine that the instrumentation at the RSD is necessary to satisfy PAM TS requirements by forgetting that RSD instruments have their own LCO. If this mistake is made, the operator could apply the basis for PAM instrumentation (part (2) of the distractor), from the basis for LCO 3.3.3.1 to this instrument. This is incorrect because RSD instrumentation has its own LCO and bases.

D. The instrument given is an instrument located on the Remote Shutdown Panel, which is required by TS LCO 3.3.3.2. The candidate could recognize this instrument as one that is required by TS, and remember the basis for this instrument, but incorrectly apply this information to the PAM Tech Spec rather than the RSD Tech Spec.

10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because it requires the SRO to have knowledge of TS Bases information required to analyze TS-required actions and terminology in that he must know the reason for an instrument to be categorized as a PAM instrument.

Reference Information:

TS 3.3.3.1 Bases 44.030.082 Accident Monitoring Reactor Vessel Water Level Division 1 Calibration/Functional NUREG 1123 KA Catalog Rev. 2 295009 Low Reactor Water Level G2.4.3 3.7/3.9 Ability to identify post-accident instrumentation 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

NRC Question Use (ILT 2020)

SRO Associated objective(s):

ILT 2020 Final Version Page: 215 of 259 Question 84 Approved View

85 K/A Importance: 4.1/4.2 Points: 1.00 S85 Difficulty: 3.00 Level of Knowledge: higher Source: BANK: PERRY 2009 91067 NRC EXAM Following a plant transient, the following conditions existed:

Performing 29.100.01 sheet 1, RPV Control, ATWS Actions.

No boron has been injected.

A short time later, the P603 Operator reports the following:

ATWS Actions are complete.

No boron has been injected.

16 control rods are NOT fully inserted.

Reactor Power is on the IRMs on Range 2 and lowering.

No boron has been injected.

Which of the following should the CRS direct?

A. Continue performing EOP sheet 1A ONLY.

B. Continue performing EOP sheet 1A AND enter 20.000.21, Reactor Scram for power control actions.

C. Exit EOP sheet 1A, AND transition RPV Level, Pressure, and Power actions to EOP sheet 1 RPV Control.

D. Exit EOP sheet 1A, transition RPV Level and Pressure actions to EOP sheet 1 RPV Control, AND enter 20.000.21, Reactor Scram, for power control actions.

Answer: B ILT 2020 Final Version Page: 216 of 259 Question 85 Approved View

Answer Explanation:

Note: Bank source for this question is the 2009 Perry NRC ILE Exam.

29.100.01 sheet 1A power leg override states: If Rx is S/D with no boron inj, then Go To: AOP 20.000.21 Per ODE-10:

ATWS Power: Rx is S/D with no boron inj means the reactor is subcritical with power below the heating range, due to control rod insertion alone. Under this condition, an exit to the scram procedure is appropriate, even though the existing margin to criticality may be small. A return to criticality is possible, but is manageable because steps in the ATWS Pressure Control leg (being performed concurrently) will terminate further cooldown, and stop the reactor power increase."

Further, since NO conditions exist that allow exit from sheet 1A, RPV control - ATWS, since there are still 16 control rods not inserted.

Therefore, the SRO applicant must determine that the reactor is shutdown with no boron injected and the power leg override noted above should be exercised, and that the current control rod position configuration indicates that remaining RPV control actions should be directed from sheet 1A because the reactor will NOT remain shutdown under all conditions w/o boron.

Distractor Explanation:

Distractors are incorrect and plausible:

A. if applicants do not recall the definition of shutdown with no boron injected and incorrectly determine all actions should remain in EOP sheet 1A. This is incorrect because the override directs power actions be taken from the scram AOP.

C. if applicants do not recall the definition of shutdown with no boron injected; and if they incorrectly believe this override provides a reason to exit sheet 1A. This is incorrect because no exit conditions exist for sheet 1A; and power actions must be directed from the scram AOP.

D. if applicants incorrectly believe the power level override provides a reason to exit sheet 1A. This is incorrect because no exit conditions exist for sheet 1A.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires that SRO applicants 1) assess plant conditions, and 2) recall the content of applicable procedures, and select the procedures with which to proceed.

The question cannot be answered solely by knowing "systems knowledge", or solely by knowing immediate operator actions, or solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs, or solely by knowing the purpose, overall sequence of events, or mitigative strategy of a procedure.

Reference Information:

29.100.01, sheets 1, 1A 20.000.21, Reactor Scram ILT 2020 Final Version Page: 217 of 259 Question 85 Approved View

NUREG 1123 KA Catalog Rev. 2 295015 Incomplete SCRAM 295015 AA2. Ability to determine and/or interpret the following as they apply to INCOMPLETE SCRAM 295015 AA2.02 4.1*/4.2* Control rod position 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level SRO Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Cognitive Terminal Given abnormal plant operating conditions and parameters, perform the required actions for the appropriate operator response in accordance with approved Fermi 2 Alarm Response Procedures, Abnormal Operating Procedures, and Emergency Operating Procedures: Provide direction and monitor the shift team during performance of an emergency / abnormal operating procedures. (SRO)

ILT 2020 Final Version Page: 218 of 259 Question 85 Approved View

86 K/A Importance: 3.8/4.0 Points: 1.00 S86 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 91827 The plant is in MODE 1, and the crew is in the process of transferring RPS A power to its alternate supply per 23.316, RPS 120V AC AND RPS MG SETS.

Prior to deenergizing RPS A, the procedure directs the operators to A. Deenergize G3352-F119, RWCU Supply Suct Iso Vlv.

Enter TS 3.6.1.3 condition A because one PCIV is inoperable in a flow path with two PCIVs.

B. Deenergize G3352-F001, RWCU Supply Inbd Iso Vlv.

Enter TS 3.6.1.3 condition A because one PCIV is inoperable in a flow path with two PCIVs.

C. Deenergize G3352-F119, RWCU Supply Suct Iso Vlv.

Enter TS 3.6.1.3 condition C because one PCIV is inoperable in a flow path with only one PCIV.

D. Deenergize G3352-F001, RWCU Supply Outbd Iso Vlv.

Enter TS 3.6.1.3 condition C because one PCIV is inoperable in a flow path with only one PCIV.

Answer: B ILT 2020 Final Version Page: 219 of 259 Question 86 Approved View

Answer Explanation:

The procedure for transferring RPS A to its alternate supply directs operators to de-energize the RWCU inboard PCIV, G3352-F001, to prevent that valve from auto closing during the power transfer. The PCIV tech spec 3.6.3.1, condition A is entered when power is removed from that valve because this flow path is one with 2 PCIVs, and is subsequently exited once power is restored after the transfer.

23.316, RPS 120V AC AND RPS MG SETS, Precautions and Limitations:

3.6 Before transferring MG Set Power Supplies, G3352-F001, RWCU Supply Inbd Iso Vlv, (G3352-F004 and G3352-F220, RWCU Supply Otbd Iso Vlv and RWCU to FW Otbd Cntm Iso Vlv), must be deenergized to prevent the valve from closing. To prevent deenergizing both valves in a single Primary Containment penetration, G3352-F001 and G3352-F004 must not be deenergized at the same time.

3.7 Anytime G3352-F001 or F004, RWCU Supply Inbd Iso Vlv or RWCU Supply Otbd Iso Vlv, or G3352-F220, RWCU to FW Otbd Cntm Iso Vlv, is de-energized while in Mode 1, 2, or 3, comply with the applicable Technical Specifications.

Section 6.1 Transferring RPS BUS A to RPS Alternate Transformer A, Step 2 states:

To prevent G3352-F001, RWCU Supply Inbd Iso Vlv, from closing:

a. De-energize G3352-F001 by placing power supply position 72B-3A Pos 5B (RB1-G13) in OPEN.

b. Comply with the applicable Technical Specifications.

Therefore, the SRO examinee must recall that F001 must be de-energized prior to transfer of power to RPS A; and recall the correct TS action to be taken with one PCIV in a flow path with 2 PCIVs.

Distractor Explanation:

A is incorrect because F119 should not be deenergized for transfer of RPS A. Plausible because this valve is deenergized for transfer of RPS B. The F119 receives an auto closure on high temperature and it is on the suction line, which is a flow path with two PCIVs.

C is incorrect because F119 should not be deenergized for transfer of RPS A. Plausible because this valve is deenergized for transfer of RPS B. The F119 receives an auto closure on high temperature and the candidate could incorrectly recall that it is in a flow path with only one PCIV.

D is incorrect. Condition C is N/A because this flow path has 2 PCIVs. Plausible if applicants do not recall the correct TS for one inop PCIV in a 2 PCIV flow path.

10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires detailed knowledge of the procedure for transferring RPS power, and the resultant tech spec evaluation necessary for application or required actions (TS Section 3).

The question cannot be answered solely by knowing </= 1-hour TS/TRM Actions, or by knowing LCO/TRM information listed "above the line" or by knowing TS safety limits.

Reference Information:

23.316, RPS 120V AC AND RPS MG SETS TS 3.6.3.1, PCIVs ILT 2020 Final Version Page: 220 of 259 Question 86 Approved View

Plant Procedures 23.316 NUREG 1123 KA Catalog Rev. 2 212000 RPS G2.1.32 Ability to explain and apply system limits and precautions 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

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

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New NRC Early Review SRO Associated objective(s):

Reactor Water Cleanup Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Identify the impact of Reactor Water Cleanup system operability on technical specifications.

ILT 2020 Final Version Page: 221 of 259 Question 86 Approved View

87 K/A Importance: 4.1/4.3 Points: 1.00 S87 Difficulty: 2.00 Level of Knowledge: higher Source: NEW 91089 The plant is at 100% power when a small steam line break occurs inside the drywell.

Drywell pressure is 2.6 psig, rising slowly.

ALL low pressure ECCS pumps are running.

HPCI injected and tripped on Level 8.

RPV level is 128, lowering slowly.

RPV Pressure is 650 psig, lowering slowly.

The crew has entered 29.100.01 sheet 1, RPV Control, and sheet 2, Primary Containment Control.

(1) What is the status of ADS timers?

(2) What action must the CRS direct?

A. (1) Timing down.

(2) Inhibit ADS.

B. (1) NOT timing down.

(2) Inhibit ADS.

C. (1) Timing down.

(2) Prevent injection from ALL low pressure ECCS pumps.

D. (1) NOT timing down.

(2) Prevent injection from ALL low pressure ECCS pumps.

Answer: D ILT 2020 Final Version Page: 222 of 259 Question 87 Approved View

Answer Explanation:

Per ARP 1D44 ADS timers will initiate when RPV level drops below L1 (31.8) AND drywell pressure exceeds 1.68# OR RPV level is 31.8 inches for 7 minutes. In this instance drywell pressure is >1.68#,

but RPV level has not dropped below L1. Therefore, ADS timers will not have initiated.

EOP sheet 2, PC Control, step PC-3 directs operators to prevent injection from CS and LPCI pumps not required for RPV injection after drywell pressure exceeds 1.68#. In this instance RPV level is adequate and high pressure injection systems are available to inject, therefore, the CRS should direct this action.

Action to inhibit ADS is NOT directed under these conditions because RPV level is NOT <L1 and the ADS timers are NOT timing down.

NOTE: This is a new procedure change due to Fermi 2 EOPs changing to match Rev 4 of the EPGs.

Prior to this change, ADS could be inhibited when it was determined that RPV Level could not be restored and maintained in the normal RPV Level Control band, thus allowing the CRS to direct inhibiting ADS before L1 was reached.

Distractor Explanation:

A is incorrect because the timers will not be timing, and ADS should not be inhibited. Plausible if applicants incorrectly believe ADS initiation signals are present, and ADS actuation must be prevented.

B is plausible if applicants incorrectly believe ADS can be inhibited at the CRS discretion, which was correct prior to Rev 4 of the EPGs. However, current guidance is to wait to inhibit ADS until initiation signals are present (less than L1 or timers counting down).

C is incorrect because ADS should not be inhibited. Plausible if applicants incorrectly believe ADS actuation must be prevented.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed.

The question cannot be answered solely by knowing "systems knowledge", or solely by knowing immediate operator actions, or solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs, or solely by knowing the purpose, overall sequence of events, or mitigative strategy of a procedure.

Reference Information:

1D44, ADS Timers Initiated 29.100.01, sheet 1, RPV Control 29.100.01, sheet 2, PC Control NUREG 1123 KA Catalog Rev. 2 218000 ADS 218000 A2. Ability to (a) predict the impacts of the following on the AUTOMATIC DEPRESSURIZATION SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

218000 A2.01 4.1/4.3* Small steam line break L0CA 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New SRO ILT 2020 Final Version Page: 223 of 259 Question 87 Approved View

Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Cognitive Terminal Given abnormal plant operating conditions and parameters, perform the required actions for the appropriate operator response in accordance with approved Fermi 2 Alarm Response Procedures, Abnormal Operating Procedures, and Emergency Operating Procedures: Provide direction and monitor the shift team during performance of an emergency / abnormal operating procedures. (SRO)

ILT 2020 Final Version Page: 224 of 259 Question 87 Approved View

88 K/A Importance: 4.6/4.6 Points: 1.00 S88 Difficulty: 3.00 Level of Knowledge: Higher Source: NEW 91107 A transient has occurred requiring emergency RPV depressurization (ED).

The crew is implementing 29.100.01, sheet 1, RPV Control.

MSIVs are open.

Feed & Condensate systems are controlling RPV water level 173-214.

The CRS directs the CRLNO to open 5 ADS SRVs.

The CRLNO reports that ONLY 4 SRVs could be opened.

RPV pressure drops to 49 psig and continues to lower slowly.

The CRS should direct the crew to A. Remotely actuate SRVs per 29.EDM.15 to continue rapid depressurization.

B. Place shutdown cooling in service using RHR pumps not required for RPV injection.

C. Defeat HPCI and RCIC isolations and align HPCI and RCIC to control RPV pressure.

D. Defeat MSIV isolations and fully open Main Turbine Bypass valves to continue rapid depressurization.

Answer: B ILT 2020 Final Version Page: 225 of 259 Question 88 Approved View

Answer Explanation:

The required ED is performed using leg S of EOP sheet 1. Step ED-5, Are 5 SRVs open?, must be answered NO. Step ED-6 then asks if RPV pressure is 52 psig. This must also be answered NO, which directs the crew to steps ED-9 and ED-10, instructing the crew to start SDC with RHR pumps not needed for injection once RPV pressure is <90 psig.

Therefore, SRO applicants must utilize plant conditions to interpret and direct execution of the appropriate EOP steps.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This action, listed in table 18 depressurization systems, is only used if RPV pressure is NOT below 52 psig after determining that <5 SRVs are open. Plausible if applicants do not recall that no additional depressurization systems should be aligned if pressure has been reduced to below 52 psig.

C. This action, listed in table 18 depressurization systems, is only used if RPV pressure is NOT below 52 psig after determining that <5 SRVs are open. Plausible if applicants do not recall that no additional depressurization systems should be aligned if pressure has been reduced to below 52 psig.

D. This action, listed in table 18 depressurization systems, is only used if RPV pressure is NOT below 52 psig after determining that <5 SRVs are open. Plausible if applicants do not recall that no additional depressurization systems should be aligned if pressure has been reduced to below 52 psig.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed.

The question cannot be answered solely by knowing "systems knowledge", or solely by knowing immediate operator actions, or solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs, or solely by knowing the purpose, overall sequence of events, or mitigative strategy of a procedure.

Reference Information:

29.100.01 sheet 1, RPV Control.

NUREG 1123 KA Catalog Rev. 2 239002 SRVs G2.1.20 Ability to interpret and execute procedure steps 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New SRO ILT 2020 Final Version Page: 226 of 259 Question 88 Approved View

Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Cognitive Terminal Given abnormal plant operating conditions and parameters, perform the required actions for the appropriate operator response in accordance with approved Fermi 2 Alarm Response Procedures, Abnormal Operating Procedures, and Emergency Operating Procedures: Monitor the shift team during performance of an emergency / abnormal operating procedure and provide feedback when not meeting expectations. (SRO)

ILT 2020 Final Version Page: 227 of 259 Question 88 Approved View

89 K/A Importance: 3.2/3.6 Points: 1.00 S89 Difficulty: 3.00 Level of Knowledge: High Source: NEW 89848 The plant is operating in MODE 1.

Operators have entered 20.300.GRID due to degraded grid conditions on the 345kV offsite electrical distribution system.

The following breakers subsequently trip open and cannot be reclosed:

Position BM - Tie Bkr Between Bus 301 and Brownstown - Enrico Fermi 3.

Position DF - Brownstown - Enrico Fermi 2 Bkr.

(1) What is the status of LCO 3.8.1, AC Sources - Operating?

(2) What electrical AOP(s), if any, would you enter if the Main Turbine Generator were to trip?

A. (1) NOT MET with one qualified offsite circuit INOPERABLE.

(2) 20.300.345kV, Loss of 345kV AOP.

B. (1) MET with all qualified offsite circuits OPERABLE.

(2) 20.300.69J, Loss of 69J and 20.300.69K, Loss of 69K AOPs.

C. (1) NOT MET with one qualified offsite circuit INOPERABLE.

(2) 20.300.69J, Loss of 69J and 20.300.69K, Loss of 69K AOPs.

D. (1) MET with all qualified offsite circuits OPERABLE.

(2) No electrical AOPs would need to be entered.

Answer: A ILT 2020 Final Version Page: 228 of 259 Question 89 Approved View

Answer Explanation:

The SRO candidate should recall the discussion of degraded offsite circuits from ODE-12, LCOs, which states: Since all analyzed accidents assume a Turbine/Generator trip, the Fermi 2 Generator cannot be relied upon to maintain incoming grid voltages. In a degraded grid voltage event with the Fermi 2 Generator on-line, ESF bus voltages may be within Tech Spec values. However, after a scram, the post trip voltages could be below the Tech Spec (UFSAR) values.

Also, TS Bases for LCO 3.8.1 states "The GDC-17 criteria are met when either BM or DF breaker is closed when the main generating unit is online. An example of not meeting the criteria is with both BM and DF breakers open, a main generator trip would open breakers CM and CF and cause a loss of the 345 kV preferred power source. Thus, an offsite circuit must be declared inoperable when the breaker alignment is such that a loss of the main generator could lead to a loss of the respective offsite circuit."

Therefore, the candidate must determine that the impact of the given breaker configuration is to declare the 345kV offsite circuit INOPERABLE because 345kV post-trip voltages would not be acceptable to sustain operability of safety-related loads as spelled out on Page 12, paragraph b, of ODE-12.

If the MTG were to trip, breakers CM and CF would trip, on reverse power, causing loss of power to Bus 301. The SRO candidate should determine that, when informed of the loss of power to Bus 301 by the CRLNO, 20.300.345kV is the correct AOP to enter as specified on Enclosure A of 30.300.345kV. NOTE:

The decision to enter specific electrical AOPs at Fermi 2 using this (and similar Enclosures) is an SRO-Only function. The RO will perform an electrical evaluation and report the results of the evaluation to the CRS, who will then take that information and decide which AOP(s) to enter.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. This combination of choices is plausible if the candidate failed to correctly recall the breaker configuration on the 345kV onsite circuit and determined that the breaker alignment in the stem of the question would cause loss of Bus 302 upon a MTG trip, thus concluding that (1) all offsite circuits are OPERABLE and (2) AOPs 20.300.69J and 20.300.69K would be entered (which would be true if only bus 302 were lost, as shown on Enclosure A of 20.300.345kV). This is incorrect because, if the MTG were to trip, breakers CM and CF would trip, on reverse power, causing loss of power to Bus 301 (not bus 302). Therefore, the 345kV offsite circuit must be declare INOPERABLE with the given breaker alignment and 20.300.345 would be entered on a trip of the MTG.

C. This combination of choices is plausible if the candidate correctly recalled the information about the breaker alignment, from TS Bases, requiring one offsite circuit to be declared INOPERABLE with BM and DF open, but didn't fully understand why, and then determined that the MTG trip will cause a loss of Bus 302, thus requiring entry into 20.300.69J and 20.300.69K. This is incorrect because, if the MTG were to trip, breakers CM and CF would trip, on reverse power, causing loss of power to Bus 301. Therefore, 20.300.345 would be entered on a trip of the MTG.

D. This combination of choices is plausible if the candidate failed to correctly recall the breaker configuration on the 345kV onsite circuit and determined that the breaker alignment in the stem of the question would not cause any power loss upon tripping of the MTG, thus concluding that (1) all offsite circuits are OPERABLE and (2) no AOP entry would be necessary post-MTG trip. This is incorrect because, if the MTG were to trip, breakers CM and CF would trip, on reverse power, causing loss of power to Bus 301. Therefore, the 345kV offsite circuit must be declare INOPERABLE with the given breaker alignment and 20.300.345 would be entered on a trip of the MTG.

10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because it cannot be answered solely by knowing 1-hour TS/TRM Actions, LCO/TRM information listed above the line, or the TS safety limits.

The question requires knowledge of TS bases that is required to analyze TS-required terminology.

Reference Information:

ODE-12, LCOs.

20.300.345kV, Loss of 345kV AOP.

20.300.GRID, Grid Disturbance AOP.

ILT 2020 Final Version Page: 229 of 259 Question 89 Approved View

SD-2500-01 System Service Diagram 4160V and 480V.

TS Bases for LCO 3.8.1.

Question Use Closed Reference ILO SRO NUREG 1123 KA Catalog Rev. 2 262001 AC Electrical Distribution 262001 A2. Ability to predict (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those abnormal conditions or operations:

262001 A2.11 3.2/3.6 Degraded system voltages 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New SRO Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Cognitive Terminal Given abnormal plant operating conditions and parameters, perform the required actions for the appropriate operator response in accordance with approved Fermi 2 Alarm Response Procedures, Abnormal Operating Procedures, and Emergency Operating Procedures: Analyze conditions and apply the appropriate technical specifications. (SRO)

ILT 2020 Final Version Page: 230 of 259 Question 89 Approved View

90 K/A Importance: 4.6 Points: 1.00 S90 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 91867 With the plant in MODE 1, EDG 11 had to run for 2 days to correct a problem with the normal feeder breaker associated with the 4160V EDG Bus that the EDG supplies. The breaker was repaired and closed and EDG 11 was shutdown.

Approximately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after shutting down EDG 11, the following data was recorded locally and provided to you, the Control Room Supervisor, for review:

EDG 11 Starting Air Receiver Pressure ------ 285 psig.

EDG 11 Fuel Oil Storage Tank Level --------- 30,000 gallons.

What is the status of these EDG parameters and what action, if any, is required?

A. Both parameters are within limits, no action is required.

B. ONLY fuel oil level is outside of allowable limits, declare EDG 11 INOPERABLE.

C. ONLY starting air receiver pressure is outside of allowable limits, declare EDG 11 INOPERABLE.

D. BOTH starting air receiver pressure AND fuel oil level are outside of allowable limits, declare EDG 11 INOPERABLE.

Answer: B ILT 2020 Final Version Page: 231 of 259 Question 90 Approved View

Answer Explanation:

Per TS Surveillance Requirement SR 3.8.3.3, EDG 11 starting air receiver pressure is within allowable limits because it is >215 psig. The SRO candidate must recall this value and determine that starting air receiver pressure is within allowable limits.

Per TS Bases for LCO 3.8.3, the fuel oil level equivalent to a 6-day supply is 30,240 gallons. Therefore, the SRO candidate must recall this value and determine that the LCO is not met, therefore Condition D is applicable requiring EDG 11 to be declared INOPERABLE.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. The SRO candidate could fail to recognize that fuel oil level is not within allowable limits, which is incorrect because fuel oil level is below the 6-day allowable value of 30,240 gallons.

C. The SRO candidate could fail to recognize that starting air pressure is within allowable limits and determine that the LCO is not met for this parameter, which is incorrect because starting air receiver pressure is >215 psig.

D. The SRO candidate could fail to recognize that starting air pressure is within allowable limits and determine that the LCO is not met for this parameter as well as for fuel oil level, which is incorrect because starting air receiver pressure is >215 psig.

10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires knowledge of TS bases that are required to analyze TS-required actions and terminology.

Although the required action to declare EDG 11 INOPERABLE is an Immediate TS Action, and therefore

<1-hour, the information necessary to get to that action is contained in the Bases for Condition A, which is SRO Only knowledge. Information regarding the lower limit for starting air pressure is in a Surveillance Requirement, which is also SRO-Only knowledge.

The question cannot be answered solely by knowing 1-hour TS/TRM Actions, or solely by knowing information in TS/TRM listed above the line or solely by knowing TS Safety Limits.

Reference Information:

Technical Specifications LCO 3.8.3, Diesel Fuel Oil and Starting Air and BASES.

NUREG 1123 KA Catalog Rev. 2 264000 Emergency Generators (Diesel/Jet)

G2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New SRO Associated objective(s):

Emergency Diesel Generator Cognitive Terminal In accordance with approved plant procedures/references, under all conditions of the Emergency Diesel Generator System: Identify the impact of the system on Technical Specifications.

ILT 2020 Final Version Page: 232 of 259 Question 90 Approved View

ILT 2020 Final Version Page: 233 of 259 Question 90 Approved View

91 K/A Importance: 3.4/3.4 Points: 1.00 S91 Difficulty: 4.00 Level of Knowledge: High Source: BANK SOURCE: 90227 FERMI 2018 NRC EXAM Reactor power is at 74% when the B3105-F031A, A RR Pump Discharge Valve, open position limit switch failed causing a loss of its full-open position indication.

(1) What is the expected plant response and (2) what shall the CRS direct?

A. (1) DCS Logic will run South RR MG Set back to the Limiter #1 setting.

(2) Direct LNO to:

Take local manual control of North RR MG Set.

Coordinate with the P603 operator to match recirc flows.

B. (1) DCS Logic will run South RR MG Set back to the Limiter #4 setting.

(2) Direct P603 to:

Verify Reactor Power < 66.1% or insert the Cram Array to lower Reactor Power to < 66.1%

Increase core monitoring for instability.

C. (1) DCS Logic will run South RR MG Set back to the Limiter #1 setting.

(2) Direct P603 to:

Verify Reactor Power < 66.1% or insert the Cram Array to lower Reactor Power to < 66.1%

Increase core monitoring for instability.

D. (1) DCS Logic will run South RR MG Set back to the Limiter #4 setting.

(2) Direct LNO to:

Take local manual control of North RR MG Set.

Coordinate with the P603 operator to match recirc flows.

Answer: B ILT 2020 Final Version Page: 234 of 259 Question 91 Approved View

Answer Explanation:

(1) RR Limiter 4 is actuated by the North (South) RR MG Set drive motor or North (South) RR MG Set generator field breakers opening. The A RR MG set breaker will trip on interlock because of the discharge valve not being full open. Once the logic is cleared (fixed), the MG set can be restarted, and the limiter must be manually reset.

(2) Actions are directed by the CRS per Condition A of 20.138.01.

Distractor Explanation:

A. Part (1) is plausible because the examinee may confuse when Limiter 1 and Limiter 4 are actuated.

RR Limiter 1 is actuated when B3105-F031A (B), N (S) RR Pump Discharge Vlv, is not fully open or total Feedwater flow is less than 20% of original rated value (approximately 3.15 Million lb./hr) for the affected RR Pump. Limiter #4 is actuated when the opposite RRMG set trips. This distractor is also plausible if the candidate failed to determine that the North RRMG would trip due to B3105-F031A going closed, causing the candidate to eliminate Limiter #4 as an option. This option is incorrect because the South RRMG set will run back to Limiter #4 upon trip of the North RRMG set. Part (2) is plausible because the candidate may not determine that the North RRMG tripped off as described above. Because of this, the candidate may determine that matching flows is required by plant procedures and TS and therefore would be a priority for the CRS. This is incorrect because matching flows is not necessary for a tripped RRMG set.

C. Part (1) is plausible because the examinee may confuse when Limiter 1 and Limiter 4 are actuated.

RR Limiter 1 is actuated when B3105-F031A (B), N (S) RR Pump Discharge Vlv, is not fully open or total Feedwater flow is less than 20% of original rated value (approximately 3.15 Million lb./hr) for the affected RR Pump. Limiter #4 is actuated when the opposite RRMG set trips. This distractor is also plausible if the candidate failed to determine that the North RRMG would trip due to B3105-F031A going closed, causing the candidate to eliminate Limiter #4 as an option. This option is incorrect because the South RRMG set will run back to Limiter #4 upon trip of the North RRMG set. Part (2) lists the correct response to be taken for a trip of the North RRMG Set.

D. Part (1) is correct, the South RRMG Set will run back to Limiter #4. Part (2) is plausible because the candidate may not determine that the North RRMG tripped off as described above. Because of this, the candidate may determine that matching flows is required by plant procedures and TS and therefore would be a priority for the CRS. This is incorrect because matching flows is not necessary for a tripped RRMG set.

Reference Information:

23.138.01 Section 1.0 (pg 4-12) 20.138.01 Condition A (pg 3) 10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because applicants must assess plant conditions and select procedural section and action to mitigate the conditions.

The question cannot be answered solely by using recirc flow control system knowledge.

ILT 2020 Final Version Page: 235 of 259 Question 91 Approved View

Plant Procedures 20.138.01 23.138.01 NUREG 1123 KA Catalog Rev. 2 202002 Recirculation Flow Control System 202002 A2. Ability to (a) predict the impacts of the following on the RECIRCULATION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

202002 A2.01 Recirculation pump trip 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILO 2018 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Higher Cognitive Level SRO Associated objective(s):

Reactor Recirculation System Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Identify the impact of Reactor Recirculation system operability on technical specifications.

ILT 2020 Final Version Page: 236 of 259 Question 91 Approved View

92 K/A Importance: 4.2 Points: 1.00 S92 Difficulty: 2.00 Level of Knowledge: High Source: NEW 91708 You are the Control Room Supervisor (CRS). A LOCA has occurred.

The plant was stabilized and Division 1 RHR placed in Torus Cooling, Torus Spray and Drywell Spray and Division 2 RHR was placed in Torus Cooling, with cooling maximized, with the following Containment parameters at the values listed:

Torus Temperature .....109°F.

Drywell Temperature ......245°F.

Torus Pressure .... 15 psig.

Drywell Pressure .. 22 psig.

The Containment parameters are now:

Torus Temperature .......93°F lowering.

Drywell Temperature ......143°F lowering.

Torus Pressure ... 0.1 psig lowering.

Drywell Pressure ... 2 psig lowering.

Which of the following will you direct?

A. Vent the Torus.

B. Vent the Drywell.

C. Terminate Torus Sprays.

D. Terminate Drywell Sprays.

Answer: C ILT 2020 Final Version Page: 237 of 259 Question 92 Approved View

Answer Explanation:

Per 29.100.01 Sheet 2, Primary Containment Control. The CRS is responsible for monitoring overrides in the EOPs and executing (directing) actions contained in the override when conditions of an override are met.

Therefore, the SRO examinee must evaluate the conditions given in the stem of the question, recall the contents of PCP-OR1, and determine that he/she should direct terminating torus spray because torus pressure is approaching 0 psig.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. PCP-OR1 contains a new override (recently added with Rev 4 of the EPGs) that requires venting primary containment (without regard to offsite rad release rates) if Primary Containment Pressure is approaching 2.0 psig. The candidate could incorrectly recall the number as being +2.0 psig or the candidate could calculate a D/P (between Torus and Drywell pressure) approaching -2.0 psig, either of which could lead the candidate to believe that he/she should continue in the PCP leg at N, which directs venting the Torus preferentially over the Drywell. This is incorrect because conditions in the stem do not indicate a need to vent Containment.

B. PCP-OR1 contains a new override (recently added with Rev 4 of the EPGs) that requires venting primary containment (without regard to offsite rad release rates) if Primary Containment Pressure is approaching 2.0 psig. The candidate could incorrectly calculate a D/P (between Torus and Drywell pressure) approaching -2.0 psig, which could lead the candidate to believe that he/she should vent the Drywell to prevent exceeding -2.0 psig from the Torus to the Drywell. This is incorrect because conditions in the stem do not indicate a need to vent Containment.

D. Conditions in the stem of the question showed that the Drywell was being sprayed for two reasons (Torus Pressure > 9.0 psig and Drywell Temperature >242°F. The candidate could evaluate existing conditions and determine that, since neither of the above are true, Drywell Sprays must be terminated. However, Drywell Sprays will be left in service, further lowering Drywell Temperature and Pressure, until DW pressure approaches 0 psig (per PCP-OR1).

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires assessment of plant conditions and then selection of a procedure section with which to proceed.

The question cannot be answered solely by knowing "systems knowledge", or solely by knowing immediate operator actions, or solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs, or solely by knowing the purpose, overall sequence of events, or mitigative strategy of a procedure.

Reference Information:

29.100.01 Sheet 2, Primary Containment Control ILT 2020 Final Version Page: 238 of 259 Question 92 Approved View

NUREG 1123 KA Catalog Rev. 2 226001 RHR/LPCI: Containment Spray System Mode G2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New SRO Associated objective(s):

Emergency and Abnormal Operating Procedures Performing Training Cognitive Terminal Given abnormal plant operating conditions and parameters, perform the required actions for the appropriate operator response in accordance with approved Fermi 2 Alarm Response Procedures, Abnormal Operating Procedures, and Emergency Operating Procedures: When multiple EOPs / AOPs are used, prioritize actions based on plant conditions and crew resources.

(SRO)

ILT 2020 Final Version Page: 239 of 259 Question 92 Approved View

93 K/A Importance: 3.6 Points: 1.00 S93 Difficulty: 3.00 Level of Knowledge: High Source: NEW 92735 The plant is in MODE 5 moving recently irradiated fuel between the RPV and the Spent Fuel Pool.

Division 1 CCHVAC is running. Division 2 CCHVAC is OPERABLE and in Auto.

A fuel handling accident then caused all Fuel Pool Ventilation Exhaust Radiation Monitors to exceed their trip setpoints.

While shifting to the Recirculation Mode, the CRLNO reports T4100-C047 Div 1 CCHVAC Emerg Makeup Fan is running with high amps and its ammeter is flashing. Everything else associated with CCHVAC and CREF has responded as expected.

(1) Which of the following actions will you direct?

(2) What is the status of CCHVAC and/or CREF OPERABILITY?

A. (1) Shut down T4100-C047.

(2) ONLY Division 1 CREF is INOPERABLE.

B. (1) Shut down T4100-C047.

(2) BOTH divisions of CREF are INOPERABLE.

C. (1) Shut down Division 1 CCHVAC and start Division 2.

(2) Only Division 1 CCHVAC is INOPERABLE.

D. (1) Shut down Division 1 CCHVAC and start Division 2.

(2) Division 1 CCHVAC and Division 1 CREF are INOPERABLE.

Answer: A ILT 2020 Final Version Page: 240 of 259 Question 93 Approved View

Answer Explanation:

ILT 2020 Final Version Page: 241 of 259 Question 93 Approved View

Per 23.413, CCHVAC System SOP, Section 7.3, CCHVAC Automatic Mode Shift to Recirculation, the ideal situation would be for both the running (Division 1 in this case) and non-running CCHVAC Emergency Makeup Fans to start in the Recirc Mode. Then, the operator would stop the makeup fan in the non-operating CCHVAC Division. However, for the conditions given in the stem, the examinee should recognize that 23.413 also allows for leaving the fan in the non-operating division running.

TS BASES for LCO 3.7.3, CREF System, states that The redundant active components of the CREF System are independent and divisionally separated. Each division operates as a unit and therefore can be considered separate subsystems, although they share non-redundant passive components. The Bases for LCO 3.7.3 state that the Redundant components, of which both divisions must be OPERABLE, include:

a. Emergency inlet air heater;

b. Emergency recirculation fans;

c. Return fans;

d. Supply fans;

e. Emergency air intakes; and

f. Air handling dampers needed to support the system operation.

Non-redundant components required to be OPERABLE include:

a. Emergency recirculation air filter train;

b. Emergency makeup air filter train; and

c. Ductwork and other system structures needed to form the necessary air flow paths.

Therefore, the examinee must determine that (1) the Division 2 makeup fan should have started (both fans receive a start signal, regardless of the running CCHVAC division, when CCHVAC shifts to Recirc) and to leave it running / shut down the Division 1 fan with the high amps; And (2) the failed makeup fan is a redundant, active component and therefore ONLY impacts Division 1 CREF OPERABILITY.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. (1) is correct. (2) is plausible is the examinee fails to recall the components that make up the redundant and non-redundant components of CREF and/or if the examinee concludes that, since BOTH emergency makeup fans start upon system initiation in the Recirc Mode, they are BOTH required for OPERABILITY. This is incorrect since either division of CREF supports either division of CCHVAC, when in the Recirc mode, and since the Emergency Makeup Fans are redundant components, of which both divisions must be OPERABLE, and therefore only one division of CREF (Div 1) is INOPERABLE.

C. (1) is plausible if the examinee determined that the emergency makeup fan associated with the running division of CCHVAC had to be running to support that division of CCHVAC and therefore concluded that the correct action is the shift running CCHVAC divisions. Part (2) is plausible if the examinee determined that the makeup fan in the stem of the question supported CCHVAC OPERABILITY and not CREF, which is plausible if the examinee hadnt studied TS Bases for CREF and CCHVAC. This is incorrect because either makeup fan can supply makeup air, through the CREF makeup filter train, regardless of which CCHVAC division is in service and the makeup fans support CREF OPERABILITY and not CCHVAC OPERABILITY.

D. (1) is plausible if the examinee determined that the emergency makeup fan associated with the running division of CCHVAC had to be running to support that division of CCHVAC and therefore concluded that the correct action is the shift running CCHVAC divisions. This line of reasoning makes (2) plausible because the examinee drawing that conclusion could plausibly determine that the failed makeup fan makes both Division 1 CREF and CCHVAC INOPERABLE. This is incorrect because either makeup fan can supply makeup air, through the CREF makeup filter train, regardless of which CCHVAC division is in service.

10 CFR 55.43(b)(2) SRO Justification:

ILT 2020 Final Version Page: 242 of 259 Question 93 Approved View

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering the question requires SRO-only specific knowledge of TS bases that is required to analyze TS-required terminology to determine OPERABILITY of the CCHVAC and CREF systems.

The question cannot be answered solely by knowing <1-hour TS/TRM Actions, LCO/TRM information listed "above the line", or by knowing TS safety limits.

Reference Information:

23.413, CCHVAC System SOP.

TS BASES for LCO 3.7.3, CREF System NUREG 1123 KA Catalog Rev. 2 290003 Control Room HVAC 290003 A2. Ability to (a) predict the impacts of the following on the CONTROL ROOM HVAC ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

290003 A2.03 3.4/3.6 Initiation/reconfiguration failure 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New SRO Associated objective(s):

Control Center HVAC Cognitive Terminal In accordance with approved plant procedures, given the condition of the system: Identify the impact of Control Center HVAC System operability on Technical Specifications.

ILT 2020 Final Version Page: 243 of 259 Question 93 Approved View

94 K/A Importance: 2.7/3.4 Points: 1.00 S94 Difficulty: 2.00 Level of Knowledge: Low Source: BANK: FERMI 2017, 90771 2012 NRC Night Orders are issued by the __(1)__ /delegate for a specified time which will not normally exceed __(2)__.

A. (1) Shift Manager (2) 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months B. (1) Director-Nuclear Operations (2) 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months C. (1) Shift Manager (2) 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months D. (1) Director-Nuclear Operations (2) 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months Answer: D ILT 2020 Final Version Page: 244 of 259 Question 94 Approved View

Answer Explanation:

Per MOP01:

3.19.1 The Director - Nuclear Operations/delegate shall issue Night Orders to be in effect for the period specified in the document. This time may be up to, but should not normally exceed, 96 hours0.00111 days 0.0267 hours 1.587302e-4 weeks 3.6528e-5 months .

A is incorrect, the SM does not have this responsibility. Plausible since SMs have many departmental procedural responsibilities.

B is incorrect, the time limit is 96 hrs. Plausible, 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is a reasonable time frame for applicants to assume if unsure.

C is incorrect, the SM does not have this responsibility; and the time limit is 96 hrs. Plausible since SMs have many departmental procedural responsibilities, and 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months is a reasonable time frame for applicants to assume if unsure.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of MOP01 it is not related to immediate actions and the entry conditions are not relevant or leading to the answer. The answer to this question is based on knowledge of administrative procedures that specify normal procedures.

Reference Information:

MOP01, section 3.19.1 Plant Procedures MOP03 - Conduct Of Operations NUREG 1123 KA Catalog Rev. 2 G2.1.15 Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, operations memos, etc.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILO 2012 Exam ILO 2017 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental SRO Associated objective(s):

Admin Procedures Exercise Cognitive Terminal Given a condition or scenario, List and describe other systems that assist Operations in the control of equipment and tracking of information in the plant, including, in accordance with the approved Fermi 2 Conduct Manuals: a. Shiftly Responsibilities; b. Audits; c. Deleted; d. Main Control Room Deficiencies and Degraded Plant Equipment; e.Â Â Â Â Â Deleted ; f. Night Orders; g. Operations Department Expectations; h. Limiting Condition for Operation; ILT 2020 Final Version Page: 245 of 259 Question 94 Approved View

95 K/A Importance: 3.9 Points: 1.00 S95 Difficulty: 2.00 Level of Knowledge: Low Source: BANK SOURCE: 2017 90149 ILT NRC EXAM Which of the following job functions is the responsibility of the Refuel Floor Supervisor during a refueling outage?

A. Maintain control of personnel entering the Refuel Floor.

B. Determine if an evacuation is necessary whenever a CAM or ARM alarms.

C. Provide the lead responsibility for development of fuel movement instructions and plans.

D. Act as Reactivity Management SRO during Control Rod testing in a cell that contains fuel.

Answer: D ILT 2020 Final Version Page: 246 of 259 Question 95 Approved View

Answer Explanation:

Per MOP13 Refuel Floor Supervisor is an SRO responsibility. MOP13 states that, during a refueling outage, the Refuel Floor Supervisor shall be on the Refuel Floor during fuel movements, and in the Control Room (as Reactivity Management SRO) during control rod testing in a cell that contains fuel.

Distractor Explanation:

All distractors are plausible because they are all functions/responsibilities performed during refueling outages by various personnel associated with the Refuel Floor and which are contained in MOP 13:

A. is a responsibility of the Refuel Floor Coordinator.

B. is a function of Radiation Protection personnel on the Refuel Floor.

D. falls under section 3.2 (Refuel Floor Supervision) of which the Refuel Floor Supervisor is part, however, this is a function of the Supervisor, Reactor Engineering.

10 CFR 55.43(b)(7) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because this question is about a refuel floor SRO responsibility that is NOT a shared responsibility with an RO Reference Information:

MOP 13, Conduct of Refueling and Core Alterations.

Plant Procedures MOP13 NUREG 1123 KA Catalog Rev. 2 G2.1.35 Knowledge of the fuel-handling responsibilities of SROs 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (7) Fuel handling facilities and procedures.

Fermi 2 NRC Exam Usage ILO 2017 Exam ILT 2020 Exam NRC Question Use (ILT 2020)

Bank Closed Reference Fundamental SRO Associated objective(s):

Admin Procedures Exercise Cognitive Terminal Given a condition or scenario, Describe methods used to control Operation s activities during refueling including, in accordance with the approved Fermi 2 Conduct Manuals: a. Refuel Floor Supervisor s responsibilities during refueling outages; b. Minimum Refuel Floor complement during core alterations; c.Â Â Â Â Â Access control for the Refuel Floor when core alterations are in progress. ; d.Â Â Â Â Â Methods for controlling refuel floor activities during non-refuel outage conditions. ;

ILT 2020 Final Version Page: 247 of 259 Question 95 Approved View

96 K/A Importance: 3.2 Points: 1.00 S96 Difficulty: 3.00 Level of Knowledge: Low Source: NEW 92148 You are the Control Room Supervisor (CRS).

A Temporary Modification (TM) package has just been brought to you in the Main Control Room.

Prior to authorizing the work package to install the TM, what are you responsible for in accordance with MES90-112, Performing Temporary Modifications?

A. Verifying the work package to install the TM is scheduled properly.

B. Establishing necessary plant conditions and ensuring compliance with Technical Specifications during installation.

C. Verifying the adequacy of the modification testing, as specified in the work package, against MMA11, Post Maintenance Testing Guidelines.

D. Verifying the work package against TM documentation and ensuring all requirements are addressed in the work package, including modification testing requirements.

Answer: B ILT 2020 Final Version Page: 248 of 259 Question 96 Approved View

Answer Explanation:

Per MES90-112 Section 6.1, Installation Instructions, Step 6.1.3 is the responsibility of the Shift Manager

/ Control Room Supervisor / Field Support Supervisor / Shift Engineer:

Establish necessary plant conditions to install the TM, ensuring compliance with Technical Specifications during installation of the TM.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This distractor is plausible because it is part of the process for installation of Temporary Modifications as per MES90-112, Section 6.1, Step 6.1.2.1. However, this distractor is incorrect because this responsibility lies with the Responsible Engineer (RE) and not the Control Room Supervisor (CRS).

C. This distractor is plausible because it is part of the process for installation of Temporary Modifications as per MES90-112, Section 6.1, Step 6.1.2.3. However, this distractor is incorrect because this responsibility lies with the Responsible Engineer (RE) and not the Control Room Supervisor (CRS).

D. This distractor is plausible because it is part of the process for installation of Temporary Modifications as per MES90-112, Section 6.1, Step 6.1.1.3. However, this distractor is incorrect because this responsibility lies with the Planner and not the Control Room Supervisor (CRS).

10 CFR 55.43(b)(3) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires specific knowledge of administrative processes for temporary modifications as outlined in MES12, Performing Temporary Modifications. To answer the question, the examinee must know actions in the Temporary Modification process that are specific to Senior Reactor Operators (SROs) only.

Reference Information:

MES90-112, Section 6.1, Installation Instructions.

Question Use Closed Reference ILO SRO NUREG 1123 KA Catalog Rev. 2 G2.2.5 2.2/3.2 Knowledge of the process for making design or operating changes to the facility 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (3) Facility licensee procedures required to obtain authority for design and operating changes in the facility.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New SRO Associated objective(s):

Admin Procedures Exercise Cognitive Terminal Given a condition or scenario, Describe who can verify installation of a Temporary Modification, in accordance with the approved Fermi 2 Conduct Manuals.

ILT 2020 Final Version Page: 249 of 259 Question 96 Approved View

97 K/A Importance: 4.3 Points: 1.00 S97 Difficulty: 3.00 Level of Knowledge: Fund Source: NEW 92267 Under which of the following conditions would the status of a system be PROVISIONALLY OPERABLE?

A. The system does not meet its Tech Spec LCO, but the system is not required to be operable in the present operating MODE.

B. The plant is in a refueling outage, the system does not meet its Tech Spec LCO, and the system is required to be operable in ALL MODES.

C. A previously inoperable system has been repaired, surveillance testing that can be performed in the present operating MODE is complete, and surveillance testing that cannot be performed without a MODE change is scheduled.

D. A previously inoperable system has been repaired, surveillance testing that can be performed in the present operating MODE is scheduled, and surveillance testing that cannot be performed without a MODE change is scheduled.

Answer: C ILT 2020 Final Version Page: 250 of 259 Question 97 Approved View

Answer Explanation:

Per MOP05 Section 4.2.3, A system is Provisionally Operable if all of the following conditions are met:

1. All previously inoperable parts of the system or component have been repaired.

2. All surveillance testing that can be performed in the present operating mode has been successfully completed.

3. The SM has completed a review of work performed and surveillance testing requirements, and verified satisfactory completion.

4. The remaining surveillance testing cannot be performed until after the plant changes operating modes (for example, systems that require steam for testing such as HPCI, RCIC, and ADS).

Distractor Explanation:

Distractors are incorrect and plausible because:

A. This is similar to the definition of Provisionally Operable and is in MOP05 (Section 4.2.2) for implementing TS Actions, but is incorrect because it is the definition of when a Tracking LCO must be written and not when to declare a system Provisionally Operable.

B. This is also in the Definition section of MOP05 (Section 4.2.1.) but is incorrect because it describes conditions that would require an Active LCO to be written, not how to determine if a system or component is Provisionally Operable.

D. The candidate could conclude that, with the system repaired and surveillances scheduled, its status could be Provisionally Operable, which is incorrect because MOP05 requires all surveillances that can be performed in the present mode to be complete prior to making that declaration.

10 CFR 55.43(b)(2) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because it involves knowledge of TS administrative requirements for determining system status as it relates to TS applicability.

The question cannot be answered solely by knowing <1-hour TS Actions, information above the line or by knowing TS safety limits.

This is an SRO Only function as described in MOP05 step 4.1.1 (Responsibility), which states: The SM/CRS/FSS is responsible for the preparation and review of Limiting Condition for Operation (LCO) sheets and Safety Function Determinations (SFDs).

Reference Information:

MOP05, Control of Equipment, Section 4.0, Implementation of Technical Specifications and TRM Actions.

NUREG 1123 KA Catalog Rev. 2 G2.2.14 Knowledge of the process for controlling equipment configuration or status.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New SRO Associated objective(s):

ILT 2020 Final Version Page: 251 of 259 Question 97 Approved View

98 K/A Importance: 3.8 Points: 1.00 S98 Difficulty: 4.00 Level of Knowledge: High Source: NEW 92787 (1) Under which of the following sets of conditions would the CRS review MES81, Control Room Envelope (CRE) Habitability Program, to determine if Potassium Iodide will be administered to Control Room Personnel?

(2) This is performed to ensure that CRE occupant radiological exposures will not exceed what value during the duration of an event?

A. (1) Any time CCHVAC is in Recirc due to High Radiation.

(2) 5 rem TEDE.

B. (1) Any time CCHVAC is in Recirc due to High Radiation.

(2) 10 rem TEDE.

C. (1) Only if CCHVAC is in Recirc, due to High Radiation, AND the CRE boundary is inoperable.

(2) 5 rem TEDE.

D. (1) Only if CCHVAC is in Recirc, due to High Radiation, AND the CRE boundary is inoperable.

(2) 10 rem TEDE Answer: C ILT 2020 Final Version Page: 252 of 259 Question 98 Approved View

Answer Explanation:

20.000.02, Abnormal Release of Radioactive Material is a generic AOP that is entered for a wide variety of radiological release events. The SRO examinee must recall that Condition I provides direction to review MES81 for requirements to administer Potassium Iodide to CR personnel when mitigating actions have been established, IAW MES81, for Control Room Envelope unfiltered air in-leakage and RBHVAC Tripped on Hi Rad or CCHVAC Recirc Mode Initiated on Hi Rad.

The SRO examinee must recall that MES81 ensures that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident, which includes the administration of KI to limit exposure when the CRE is inoperable. Taking KI saturates the thyroid gland with stable (non-radioactive) iodine. This prevents or reduces the amount of radioiodine that will be taken up by the thyroid.

Distractor Explanation:

Distractors are incorrect and plausible because:

Part (1) is plausible because any time CCHVAC is in Recirc, due to High Rad conditions, personnel in the CRE have the potential to receive exposure to ionizing radiation in excess of normal values. However, with the CRE boundary operable, evaluation of MES81 requirements for administration of KI is not necessary because operability of the CRE boundary ensures that the in-leakage of unfiltered air into the CRE will not exceed the in-leakage assumed in the licensing basis analysis of design basis accident (DBA) consequences to CRE occupants (from TS Bases for LCO 3.7.3).

Part (2) is plausible because 10 rem TEDE is the Whole Body dose limit in EP-201-03, Variances from Routine Radiological Practice and Procedures During an Emergency for mitigating an accident. Since Control Room Operators occupy the CRE for the purpose of mitigating an accident, it is plausible that an examinee could relate this dose limit with the basis for actions taken in MES81. This is incorrect because MES81 states that the Control Room Envelope Habitability Program ensures that adequate radiation protection is provided to permit access and occupancy of the CRE under design basis accident (DBA) conditions without personnel receiving radiation exposures in excess of 5 rem total effective dose equivalent (TEDE) for the duration of the accident 10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because answering this question requires the SRO examinee to assess plant conditions and then select the procedure section with which to proceed.

The question cannot be answered solely by knowing "systems knowledge", or solely by knowing immediate operator actions, or solely by knowing entry conditions for AOPs or plant parameters that require direct entry into major EOPs, or solely by knowing the purpose, overall sequence of events, or mitigative strategy of a procedure.

Reference Information:

20.000.02, Abnormal Release of Radioactive Material.

MES81, Control Room Envelope (CRE) Habitability Program.

EP-201-03, Variances from Routine Radiological Practice and Procedures During an Emergency.

LCO 3.7.3 BASES.

ILT 2020 Final Version Page: 253 of 259 Question 98 Approved View

NUREG 1123 KA Catalog Rev. 2 G2.3.14 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Higher Cognitive Level New SRO Associated objective(s):

RERP: Emergency Classifications and Protective Action Recommendations Performance Enabler Ensure that potassium iodine (KI) is administered in accordance with instructions provided by the DTE Energy Medical Director or authorized representative ILT 2020 Final Version Page: 254 of 259 Question 98 Approved View

99 K/A Importance: 4.4 Points: 1.00 S99 Difficulty: 2.00 Level of Knowledge: Fund Source: NEW 92200 The EOPs have been entered from 100% power.

Adequate Core Cooling is in question.

The STA reports that CHRRMs is currently reading 6,570 R/hr.

Emergency Response facilities are not yet functional.

Which of the following is correct regarding entry into the SAGs?

A. Enter the SAGs IF CHRRMs rises to >10,000 R/hr.

B. Current conditions indicate that SAG entry is required NOW.

C. Enter the SAGs WHEN notified that the Emergency Response facilities are functional.

D. Do not enter the SAGs UNTIL the determination that core damage is occurring is made by personnel in the Emergency Response facilities.

Answer: A ILT 2020 Final Version Page: 255 of 259 Question 99 Approved View

Answer Explanation:

Per ODE-10, Emergency Operating Procedure Expectations:

CORE DAMAGE means widespread degradation of the fuel pellet or cladding fission product barriers due to inadequate core cooling. Localized fuel rod failures that may occur during normal power operation as a consequence of pellet-clad interaction, hydriding, fretting damage, crud-induced corrosion, etc., are not considered to be core damage. The determination that CORE DAMAGE is occurring (requiring SAG entry) will usually be made by personnel in the emergency response facilities (TSC or EOF). If information from the emergency facilities is not yet available, a CHRRMs reading of >10,000 R/hr is considered indication that CORE DAMAGE is occurring, basis for the correct answer With the emergency response facilities currently unavailable, SAG entry would be made when CHRRMs rises to >10,000 R/hr, because this would be considered Core Damage and thus meet the requirements of various EOP overrides to enter the SAGs.

Distractor Explanation:

Distractors are incorrect and plausible because:

B. CHRRMs is significantly elevated and above 2250 R/hr, which is the value in EP-101, Emergency Classifications, that the Fission Product Barrier is classified as LOST. The candidate could put the questionable adequate core cooling with this information and determine that SAG entry is required.

This is incorrect because, with the emergency response facilities unavailable, ODE-10 puts the threshold for SAG entry at 10,000 R/hr on CHRRMs .

C. CHRRMs is significantly elevated and above 2250 R/hr, which is the value in EP-101, Emergency Classifications, that the Fission Product Barrier is classified as LOST. The candidate could put the questionable adequate core cooling with this information and determine that SAG entry will be required, but only when the emergency response facilities become available. This is incorrect because ODE-10 puts the threshold for SAG entry at 10,000 R/hr on CHRRMs (when emergency response facility information is unavailable). The emergency response facilities becoming functional does not change this.

D. The candidate could conclude that SAG determination is not made without information from the emergency response facilities. This is not correct since ODE-10 provides guidance for making the determination to enter the SAGs if the emergency response facilities are unavailable.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because it involves assessment of plant conditions and decision points in the EOPs that involve transition to emergency contingency procedures.

The question cannot be answered solely by knowing systems knowledge, immediate operator actions, EOP entry conditions or purpose, overall sequence of events, or overall mitigative strategy of a procedure.

Reference Information:

ODE-10, Emergency Operating Procedure Expectations.

EP-101, Emergency Classifications.

ILT 2020 Final Version Page: 256 of 259 Question 99 Approved View

NUREG 1123 KA Catalog Rev. 2 G2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as operating procedures, abnormal operating procedures, and severe accident management guidelines 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Closed Reference Fundamental New SRO Associated objective(s):

Severe Accident Guidelines - In-Depth Cognitive Terminal Summarize the transition into SAGs and how SAGs supplement the EOPs.

ILT 2020 Final Version Page: 257 of 259 Question 99 Approved View

100 K/A Importance: 4.1 Points: 1.00 S100 Difficulty: 2.00 Level of Knowledge: High Source: MODIFIED 92160 The plant is in MODE 1 at 100% power.

The Outside Rounds NO reports that there is an on-going gun fight with multiple intruders and Security officers by the GSW Intake Structure traveling screen.

Security confirms this report and adds that there has been an explosion in a manhole adjacent to the GSW Pump House.

Outside Rounds also reports he is exiting the area and there is no visible damage to the Service Water system.

What is the appropriate emergency classification for this event?

A. Unusual Event.

B. Alert.

C. Site Area Emergency.

D. General Emergency.

Answer: B ILT 2020 Final Version Page: 258 of 259 Question 100 Approved View

Answer Explanation:

NOTE: EP-101 Event Classification charts to be provided for this question.

NOTE: This question is modified from Question S99 on the 2017 ANO NRC exam. The question was modified by changing the correct answer, from SAE to Alert, because the intake at Fermi 2 is located outside of the Protected Area.

Per EP-101, the appropriate emergency classification is Alert, based on a HOSTILE ACTION occurring within the OWNER CONTROLLED AREA as reported by the Security Shift Supervisor under EAL HA1.1.

Distractor Explanation:

Distractors are incorrect and plausible because:

A. Unusual Event is the correct classification for a SECURITY CONDITION that does NOT involve a HOSTILE ACTION per EAL HU1.1. This classification is incorrect because the stem indicates the threat is a HOSTILE ACTION inside the Owner Controlled Area.

C. SAE is the correct classification for a HOSTILE ACTION with the PROTECTED AREA per EAL HS1.1. This classification is incorrect because the stem indicates that the HOSTILE ACTION is inside the Owner Controlled Area, but not the Protected Area.

D. GE is the correct classification for a HOSTILE ACTION within the PROTECTED AREA that causes loss of a critical safety function per EAL HG1.1. This classification is incorrect because, although the stem indicates that an explosion has occurred, it did not threaten any of the listed safety functions nor did it cause damage to spent fuel.

10 CFR 55.43(b)(5) SRO Justification:

This question meets ES-401 Attachment 2 requirements to be SRO-Only because it requires assessment of plant conditions (normal, abnormal, or emergency) and then selection of a procedure or section of a procedure to mitigate or recover, or with which to proceed. It cannot be answered solely by knowing system knowledge, immediate actions, EOP/AOP entry conditions or the overall mitigating strategy of procedures.

Reference Information:

EP-101, Classification of Emergencies.

NUREG 1123 KA Catalog Rev. 2 G2.4.28 Knowledge of procedures relating to a security event (non-safeguards information) 10CFR55 RO/SRO Written Exam Content 10 CFR 55.43(b) (5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Fermi 2 NRC Exam Usage ILT 2020 Exam NRC Question Use (ILT 2020)

Higher Cognitive Level Modified Reference Provided SRO Associated objective(s):

RERP: Emergency Classifications and Protective Action Recommendations Performance Enabler Classify Emergency Events ILT 2020 Final Version Page: 259 of 259 Question 100 Approved View

RO Reference Handouts

20.000.19 Attachment 1, Page 1 of 3 021601 REACTOR COOLDOWN MONITORING SATURATED STEAM TABLES Gage Pressure Temperature Gage Pressure Temperature Lb./Sq. In. ºF Lb./Sq. In. ºF 0 213º 200 388º 5 228º 210 392º 10 239º 220 396º 15 250º 230 399º 20 259º 240 403º 25 267º 250 406º 30 274º 260 409º 35 281º 270 413º 40 287º 280 416º 45 293º 290 419º 50 298º 300 422º 55 303º 350 436º 60 307º 400 448º 65 312º 450 459º 70 316º 500 470º 75 320º 550 480º 80 324º 600 489º 85 328º 650 497º 90 331º 700 505º 95 335º 750 513º 100 338º 800 520º 105 341º 850 527º 110 344º 900 534º 115 347º 950 540º 120 350º 1000 546º 125 353º 1050 552º 130 356º 1100 558º 135 358º 1150 563º 140 361º 1200 569º 145 364º 1250 574º 150 366º 1300 579º 160 371º 1350 584º 170 375º 1400 588º 180 380º 1450 593º 190 384º 1500 598º

Detroit Edison - Fermi 2 EP-201-03 RERP Plan Implementing Procedure Revision 9 Page 1 VARIANCES FROM ROUTINE RADIOLOGICAL PRACTICE AND PROCEDURES DURING AN EMERGENCY Revision Summary

1) Deleted Attachments 1 and 2 from procedure and recreated as forms.

2) Added step 2.2 to reference Appendix A for new location of forms.

3) Deleted previously embedded commitments in Section 6.0 that no longer apply.

4) Made editorial changes throughout procedure.

Implementation Plan

1) This revision goes into effect upon approval.

Enclosures - None Information and Procedures DTC DSN Revision Date Issued DCR # File # IP Code: Recipient TPEPT EP-201-03 9 10/7/2009 09-1269 1703.10 I

EP-201-03 Revision 9 Page 2 1.0 PURPOSE To describe variances from routine radiological practices and procedures, for use with proper authorization, in the event of a radiological emergency.

2.0 USE REFERENCES 2.1 EP-226, Potassium Iodide 2.2 Appendix A, RERP Forms 3.0 ENTRY CONDITIONS 3.1 The variances discussed in this procedure may be implemented during a declared emergency event as directed by the Emergency Director or, alternately, at the discretion of Radiation Protection (RP) personnel. In some cases, these variances may be deemed appropriate by the individual at the scene, if adherence to routine practices and procedures would adversely impact the emergency response (e.g., accident mitigation or first aid/life saving activities).

3.2 In general, the following criteria can be applied to an emergency event to determine the applicability of varying from routine procedures:

3.2.1 The event involves an actual or potential threat to personnel, the plant, or the public.

3.2.2 The event requires a timely (or immediate) response.

4.0 GENERAL INFORMATION 4.1 Authorized variances from routine radiological practices and procedures include a relaxation or reduction of some administrative radiological controls. This is done for the purpose of making a timely (or immediate) response to emergency conditions.

4.2 Generally, these variances represent a shift from a program of long-term protection against occupational exposure to a program of immediate safety from high radiation exposures.

What remains unchanged in both cases is that all exposures must be planned and justified.

They must also be controlled and monitored to the extent practical.

EP-201-03 Revision 9 Page 3 4.3 The types of variances which may be authorized during an emergency include:

4.3.1 Increase in personnel exposure limits 4.3.2 Decrease in requirements relating to Radiation Work Permits (RWPs) and access control 4.3.3 Reduction in equipment requirements (protective clothing and respirators) 5.0 IMMEDIATE ACTIONS 5.1 Emergency Director NOTE: Authorization to exceed regulatory limits can only be granted by the Emergency Director. This responsibility cannot be delegated.

5.1.1 May authorize emergency workers to exceed regulatory limits stated in step 6.1.1.2.

5.1.2 Ensures all attempts to maintain personnel exposure within regulatory limits have been attempted or evaluated and failed or determined unfeasible.

5.1.3 Ensures all exposures are planned.

5.1.4 Ensures all personnel performing emergency activities involving exposures which may exceed regulatory limits are volunteers and shall be briefed on potential exposure consequences prior to receiving such exposure.

5.1.5 May authorize personnel to enter areas for short-time surveillance or response functions without the equipment normally required. This may include:

1. Entering a contaminated area without protective clothing

2. Entering an airborne radioactivity area without respiratory equipment

3. Entering a high radiation area without a high range direct-reading or alarming personnel dosimeter.

EP-201-03 Revision 9 Page 4 5.2 Radiation Protection Personnel NOTE: RWPs that are verbally issued must be later prepared by Radiation Protection personnel when time allows.

5.2.1 May suspend the requirement for written, approved RWPs prior to commencing work and verbally issue an RWP to allow for timely emergency response and rapid changes in working conditions.

5.2.2 May authorize personnel to violate or disregard radiological barriers or signs for the performance of a particular job or function.

5.2.3 May issue special and/or emergency dosimetry to responders. All issuance will be documented on EP-201031 and/or EP-201032 as appropriate.

5.3 Self Monitors NOTE: The increased or changing priorities for Radiation Protection during an emergency may limit the immediate availability of RP personnel for performance of surveillance and monitoring. Therefore, personnel who have been trained and are qualified in self monitoring techniques may be utilized for support.

5.3.1 Perform surveillance and monitoring of their own work 5.3.2 Supplement plant-wide surveillance capabilities 5.4 Plant Personnel:

5.4.1 May disregard RWP and RP access control restraints (i.e., ropes, barriers, etc.)

without authorization from RP for the purposes of accident mitigation, rescue, and rendering of major first aid if they:

1. Have a means of determining radiation levels (i.e., indication of a nearby area radiation monitor, available survey instrument, or digital alarming dosimeter.

2. Have a means of monitoring their own exposure (DRD).

3. Remain within regulatory exposure limits.

EP-201-03 Revision 9 Page 5 6.0 PROCEDURE NOTE: Emergency exposures stated below should be limited to once in a lifetime.

6.1 Emergency Extensions Exceeding Fermi 2 Administrative Dose Limits 6.1.1 Radiation Protection Advisor or Emergency Director:

1. Determine emergency conditions exist that warrant exceeding administrative dose limits up to the following federal occupational dose limits.

NOTE: Projected or actual thyroid exposures to radioiodine 25 Rem CDE should be minimized by the use of Potassium Iodide (KI) and/or respirators in accordance with EP-226.

2. May authorize dose extensions up to the federal occupational limits (Table 1) provided emergency workers:

a. Are adequately trained on appropriate Radiation Protection procedures.

b. Have a record of their year-to-date dose.

c. Are fully briefed regarding their duties and expected actions, actual or expected radiological conditions, stay times and other hazards.

d. Maintain exposures ALARA.

Table 1 Federal Occupational Dose Limits (Annual) 5 Rem TEDE Whole Body 15 Rem LDE Lens of the Eye 50 REM SDE Skin/Extremities 50 REM TODE Organs (i.e., Thyroid)

EP-201-03 Revision 9 Page 6 6.2 Emergency Extensions Exceeding Federal Occupational Dose Limits NOTE: Authorization to exceed federal occupational dose limits can only be granted by the Emergency Director. Emergency exposures involve non-pregnant adults, and to the extent practical, should be limited to 5 Rem TEDE and once in a lifetime.

6.2.1 Emergency Director:

1. Determine emergency conditions exist that warrant exceeding federal occupational dose limits.

2. May authorize exceeding federal occupational dose limits provided that all emergency workers meet the criteria stated in Section 6.1.1.2.

3. May authorize the following dose limits to mitigate an accident or protect valuable property:

Table 2 Dose Limits to Mitigate an Accident or Protect Valuable Property 10 Rem TEDE Whole Body 30 Rem LDE Lens of the Eye 100 REM SDE Skin/Extremities 100 REM TODE Organs (i.e., Thyroid)

4. May authorize the following dose limits to save lives or protect large populations:

Table 3 Dose Limits to Save Lives or Protect Large Populations 25 Rem TEDE Whole Body 75 Rem LDE Lens of the Eye 250 REM SDE Skin/Extremities 250 REM TODE Organs (i.e., Thyroid)

EP-201-03 Revision 9 Page 7

5. May authorize whole body exposures in excess of 25 Rem to save lives or protect large populations provided all emergency workers:

a. Meet criteria listed in step 6.1.1.2.

b. Are volunteers.

c. Are made fully aware of the risks listed in Tables 4 and 5.

Table 4 Health Effects Associated with Whole-Body Absorbed Doses Received Within a Few Hours a Whole Whole Prodromal Body Early Body Effects c Absorbed Fatalities b Absorbed (Percent Dose (Rad) (Percent) Dose (Rad) Affected) 140 5 50 2 200 15 100 15 300 50 150 50 400 85 200 85 460 95 250 98 a

Risks will be lower for protracted exposure periods.

b Supportive medical treatment may increase the dose at which these frequencies occur by approximately 50 percent.

c Forewarning symptoms of more serious health effects associated with large doses of radiation.

EP-201-03 Revision 9 Page 8 Table 5 Approximate Cancer Risk to Average Individuals from 25 Rem Effective Dose Equivalent Delivered Promptly Age Appropriate Risk of Average Years of Life at Premature Death Lost if Premature Exposure (Deaths per 1000 Death Occurs (Years) Persons Exposed) (Years) 20 to 30 9.1 24 30 to 40 7.2 19 40 to 50 5.3 15 50 to 60 3.5 11 7.0 FOLLOW UP ACTIONS 7.1 Completed forms shall be forwarded to the Supervisor, RERP for disposition.

8.0 RECORDS 8.1 The following are required records and shall be retained or dispositioned in accordance with established requirements:

8.1.1 Emergency Dosimetry Issue Log (EP-201031) 8.1.2 Special Dosimetry Issue Log (EP-201032)

END

ARP 3D19 Rev 20 Reference Use Page 1 ANNUNCIATOR SYSTEM TROUBLE AUTO ACTIONS NOTE: The system that had the critical failure will not automatically restart.

1. Failure of any software module designated as critical will cause an automatic failover of the system.

NOTE: A total loss of VAS indication on H11-P603 OR a loss of data acquisition MUX A and C may meet the loss of monitoring criteria for EP-101 EAL SU3.1, or SA3.1.

INITIAL RESPONSE NOTE: Visual Annunciator System (VAS) Multiplexer (MUX) failure may occur that results in a continuous rebooting of the MUX or false alarms in the Control Room. In either case the MUX not working properly needs to be identified promptly. Upon receipt of alarm 3D19, ANNUNCIATOR SYSTEM TROUBLE, prompt identification of the associated MUX source can be performed using VAS screens as described.

1. Upon receipt of 3D19, ANNUNCIATOR SYSTEM TROUBLE, alarm MUX related events will be noted by:

At least 2 SOERs will be generated VAS HARDWARE SYSTEM TROUBLE (848C97), VAS SOFTWARE SYSTEM TROUBLE (849C97).

Alarm Screen (TOC=ALARM) will display at the top part (VAS SYSTEM) specific alarm as to which MUX failed.

Alarm Screen (TOC=ALARM) will display at the bottom part (INPUT HEALTH) 1280 points in alarm (UNHEALTHY).

Annunciator Status screen (TOC=STATUS) will show the status of the failed MUX (realtime)

VAS System Software Trouble Overview (TOC=SWTRBL) will show the status for the failed MUX processor (realtime)

VAS System Hardware Trouble Overview (TOC=HWTRBL) will show the status of the Failed MUX and its associated Watchdog Timer Card status.

If the initiating condition is a MUX failure and the system does not auto recover or fails again and does not recover within 15 minutes and Action A fails to recover the MUX, remove the failed MUX from service in accordance with 23.621, "Main Control Room Annunciator And Sequence Recorder."

Information and Procedures DTC DSN Revision Date Issued DCR # File # IP Code: Recipient TPARP 3D19 20 04/07/2017 17-0549 1703.02 I

ARP 3D19 Rev 20 Page 2 NOTE: If the following steps do not correct the problem contact I&C / System Engineer for assistance.

2. Review the Annunciator System Trouble Overview (TOC=TROUBLE), Hardware System Trouble (TOC=HWTRBL), and Software System Trouble (TOC=SWTRBL) displays to determine system status. The attached Flow Chart may assist with Alarm diagnosis.

3. IF the initiating condition is Ground Fault Detection, perform 23.621 Attachment 4, Ground Isolation Procedure for VAS.

4. IF the initiating condition is a critical failure of a software module, restart the computer that automatically shutdown per 23.621, "Main Control Room Annunciator And Sequence Recorder."

5. IF the initiating condition is a non-critical failure of any software module, perform the following:

If the system is operating in the prime/back mode, perform a system failover per 23.621, "Main Control Room Annunciator And Sequence Recorder."

Verify software module restarted and is running properly on the new prime computer.

6. IF the initiating condition is a high temperature on any H11-P827 cabinet perform the following:

Open the cabinet doors (both front and rear)

Close or verify closed circuit breaker RL01-CB1.2 in the effected cabinet.

Verify the cabinet cooling fan is operating

7. IF the initiating condition is a Watchdog Timer Card perform the following:

Verify all circuit breakers in the effected cabinet are closed.

Verify all chassis in the effected cabinet are turned on

Verify at the affected MUX (G2) that power is on and the green light above the network connection is lit.

8. IF the initiating condition is a chassis tickle failure, reset the tickle lock-in per 23.621, "Main Control Room Annunciator And Sequence Recorder."

ARP 3D19 Rev 20 Page 3 SUBSEQUENT ACTIONS

1. Document all conditions and actions on a CARD for tracking or correction.

INITIATING DEVICE/SETPOINTS Ground Fault Detectors:

C97K051 C97K056 C97K052 C97K057 C97K053 C97K058 C97K054 C97K059 C97K055 C97K060 Cabinet High Temperature Detectors:

C97K062 (A1-HB01) C97K132 (A8-HB01)

C97K070 (A2-HB01) C97K078 (A9-HB01)

C97K110 (A3-HB01) C97K086 (A10-HB01)

Internally generated see displays Hardware Trouble (TOC=HWTRBL) and Software Trouble (TOC=SWTRBL)

REFERENCES I-2080-16 I-2080-48 I-2080-71 I-2080-72 I-2080-79 VAS User Manual

ARP 3D19 Rev 20 Page 4

System running ANNUNCIATOR Primary CPU only Failure of any Critical

HW Trouble SYSTEM Auto Actions Software Module will SOER in alarm TROUBLE cause an automatic failover

Hardware System Trouble (HWTRBL)

Review Annunciator

Software System System Trouble Trouble (SWTRBL) System Running Restart Overview

SOER Screen (SOER) properly with Primary Y Backup (TROUBLE) only ? CPU N

Initiating Event Analysis H11-P603 Backup TOTAL Y Loss of VAS CPU Y LOSS OF VAS Alarm ? available ?

Contact N Ground Fault Y Maintenance - I&C Detection ?

for ground isolation Initiate N

CARD N

H21-P827 Cabinet Open Y

Temperatures ? Cabinet Doors Close or Verify Closed affected cabinet Circuit Breaker N

Annunciator Verify cabinet System cooling fan is Trouble operating Alarm Clears Hardware System Trouble ?

Y 2 N

Software System Trouble Y 4 TERMINATE

ARP 3D19 Rev 20 Page 5 2

Hardware 3

System Trouble Chassis Watchdog Process Nodes /

Tickle Card Timer Network Reset the chassis tickle Associated Watchdog CPU x status = DOWN latches per 23.621 Y Chassis Power ON ? CPU y status = PRIMARY Y N Wait for the Cycle power on chassis tickle alarmed watchdog Network SWCCRB frequency time timer card chassis Y alarming ?

N N

Watchdog card alarm clears ? MCR SDS PCFMCR alarming Y

Chassis tickle N Use a different alarm 3D19 MCR SDS workstation clear ? for VAS functions Contact Maintenance - I&C for assistance Y TERMINATE

ARP 3D19 Rev 20 Page 6 3

Hardware System Trouble Have a number of DATA ACQUISITION N False alarm windows Y STATUS turned on?

POINT ID STATUS STATUS STATUS STATUS Identify MUX MUX-A FAILED PRIMARY PRIMARY PRIMARY Responsible MUX-C PRIMARY FAILED PRIMARY PRIMARY MUX-B PRIMARY PRIMARY FAILED PRIMARY MUX-D PRIMARY PRIMARY PRIMARY FAILED Remove MUX from service iaw 23.621 Section 7.13 ACTION A

MUX-A FAILED PRIMARY PRIMARY FAILED MUX-C PRIMARY FAILED FAILED PRIMARY Initiate CARD for MUX-B PRIMARY FAILED PRIMARY FAILED workorder MUX-D FAILED PRIMARY FAILED PRIMARY ACTION B

MUX-A FAILED PRIMARY TERMINATE MUX-C FAILED PRIMARY MUX-B PRIMARY FAILED MUX-D PRIMARY FAILED ACTION C

MUX-A FAILED FAILED PRIMARY FAILED MUX-C FAILED FAILED FAILED PRIMARY MUX-B PRIMARY FAILED FAILED FAILED MUX-D FAILED PRIMARY FAILED FAILED ACTION D

MAIN PROCESSING / DATA ACQUISITION

ARP 3D19 Rev 20 Page 7

ARP 3D19 Rev 20 Page 8 END

SRO Reference Handouts

EP-101 Enclosure A, Page 1 of 3 101819A EMERGENCY ACTION LEVEL (EAL) CLASSIFICATION MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous radioactivity resulting in offsite dose greater Release of gaseous or liquid radioactivity resulting in offsite Release of gaseous or liquid radioactivity greater than 2 times than 1,000 mrem TEDE or 5,000 mrem thyroid CDE than 100 mrem TEDE or 500 mrem thyroid CDE dose greater than 10 mrem TEDE or 50 mrem thyroid CDE the ODCM limits for 60 minutes or longer RG1.1 1 2 3 4 5 DEF RS1.1 1 2 3 4 5 DEF RA1.1 1 2 3 4 5 DEF RU1.1 1 2 3 4 5 DEF In the absence of real-time dose assessment, reading on In the absence of real-time dose assessment, reading on In the absence of real-time dose assessment, reading Reading on any Table R-1 effluent radiation monitor any Table R-1 effluent radiation monitor > column "ALERT" any Table R-1 effluent radiation monitor > column "GE" for on any Table R-1 effluent radiation monitor > column > column "UE" for 60 min. (Notes 1, 2, 3) for 15 min. (Notes 1, 2, 3, 4) 15 min. (Notes 1, 2, 3, 4) "SAE" for 15 min. (Notes 1, 2, 3, 4)

RG1.2 1 2 3 4 5 DEF RS1.2 1 2 3 4 5 DEF RA1.2 1 2 3 4 5 DEF RU1.2 1 2 3 4 5 DEF Dose assessment using actual meteorology indicates Dose assessment using actual meteorology indicates Dose assessment using actual meteorology indicates doses Sample analyses for a gaseous or liquid release indicates doses > 1,000 mrem TEDE or 5000 mrem thyroid CDE at doses > 100 mrem TEDE or 500 mrem thyroid CDE at or > 10 mrem TEDE or 50 mrem thyroid CDE at or beyond the a concentration or release rate 2 x ODCM limits for 60 or beyond the SITE BOUNDARY (Notes 3, 4) beyond the SITE BOUNDARY (Notes 3, 4) SITE BOUNDARY (Notes 3, 4) min. (Notes 1, 2) 1 RG1.3 1 2 3 4 5 DEF RS1.3 1 2 3 4 5 DEF RA1.3 1 2 3 4 5 DEF Rad Field survey results indicate EITHER of the following at or Field survey results indicate EITHER of the following at or Analysis of a liquid effluent sample indicates a concentration Effluent beyond the SITE BOUNDARY: beyond the SITE BOUNDARY: or release rate that would result in doses > 10 mrem TEDE Closed window dose rates > 1,000 mR/hr expected Closed window dose rates > 100 mR/hr expected to or 50 mrem thyroid CDE at or beyond the SITE BOUNDARY to continue for 60 min. continue for 60 min. for 60 min. of exposure (Notes 1, 2)

Analyses of field survey samples indicate thyroid Analyses of field survey samples indicate thyroid 1 2 3 4 5 DEF RA1.4 CDE > 5,000 mrem for 60 min. of inhalation. CDE > 500 mrem for 60 min. of inhalation.

(Notes 1, 2) (Notes 1, 2) Field survey results indicate EITHER of the following at or beyond the SITE BOUNDARY:

Closed window dose rates > 10 mR/hr expected to continue for 60 min.

Analyses of field survey samples indicate thyroid CDE > 50 mrem for 60 min. of inhalation.

R (Notes 1, 2)

Abnormal Spent fuel pool level cannot be restored to at least the top of the Spent fuel pool level at the top of the fuel racks Significant lowering of water level above or damage to irradiated Unplanned loss of water level above irradiated fuel Rad fuel racks for 60 minutes or longer fuel Levels RG2.1 1 2 3 4 5 DEF RS2.1 1 2 3 4 5 DEF RA2.1 1 2 3 4 5 DEF RU2.1 1 2 3 4 5 DEF

/

Rad Spent fuel pool level cannot be restored to at least Level 3 Lowering of spent fuel pool level to Level 3 as indicated Uncovery of irradiated fuel in the REFUELING PATHWAY UNPLANNED water level drop in the REFUELING PATHWAY Effluent as indicated by level > 15 ft. on G41R601A/B for by level < 15 ft. on G41R601A/B as indicated by any of the following:

RA2.2 1 2 3 4 5 DEF

> 60 min. (Note 1) 2D1, FUEL POOL WATER LEVEL LOW alarm 2 Table R-1 Effluent Monitor Classification Thresholds Damage to irradiated fuel resulting in a release of radioactivity AND Floodup Level Transmitter (when in service)

Visual observation Irradiated Any of the following radiation monitor indications: AND Fuel Event Release Point Monitor GE SAE Alert UE UNPLANNED rise in area radiation levels as indicated by any RB5 Refuel Floor Hi Range ARM (Ch. 18) alarm RB Ventilation SPING (Ch. 5) N/A N/A N/A 5.2E-3 µCi/cc RBHVAC Vent Exhaust Radiation Monitor > 16,000 cpm of the following radiation monitors:

Fuel Pool Vent Exhaust Radiation Monitor > 5 mR/hr RB5 Spent Fuel Pool ARM (Ch. 15)

SPING (Ch. 7) N/A N/A N/A 1.5E-1 µCi/cc RB5 Refuel Floor Lo Range ARM (Ch. 17)

SGTS Div. I RA2.3 1 2 3 4 5 DEF RB5 Refuel Floor Hi Range ARM (Ch. 18)

AXM (Ch. 3/4) 2.9E+3 µCi/cc 2.9E+2 µCi/cc 2.9E+1 µCi/cc N/A Lowering of spent fuel pool level to Level 2 as indicated by Gaseous SPING (Ch. 7) 1.4E-1 µCi/cc level < 33 ft. on G41R601A/B N/A N/A N/A SGTS Div. II Radiation levels that impede access to equipment necessary for AXM (Ch. 3/4) 2.8E+3 µCi/cc 2.8E+2 µCi/cc 2.8E+1 µCi/cc N/A normal plant operations, cooldown or shutdown RW Ventilation SPING (Ch 5) N/A N/A N/A 3.5E-2 µCi/cc 3 TB Ventilation None SPING (Ch 5) N/A N/A N/A 3.1E-4 µCi/cc RA3.1 1 2 3 4 5 Dose rate > 15 mR/hr in EITHER of the following areas:

DEF Area AB3 Control Room (ARM Channel 6)

Radiation Liquid Levels CW RSVR Central Alarm Station (by survey)

D11-R806 N/A N/A 1.1E+6 cpm 1.3E+4 cpm Decant Line RA3.2 1 2 3 4 5 DEF An UNPLANNED event results in radiation levels that prohibit or impede access to any Table R-2 rooms or areas (Note 5)

Table R-2 Safe Shutdown Rooms/Areas Damage to a loaded cask CONFINEMENT BOUNDARY Room/Area Mode Applicability EU1.1 1 2 3 4 5 DEF E 1 None Relay Room AB3-DC MCC Area All 3 None Damage to a loaded cask CONFINEMENT BOUNDARY as indicated by an on-contact radiation reading greater than EITHER of the following on the surface of the spent fuel cask Confinement ISFSI RB1-F17 3 (overpack):

Boundary RB1-F11 3 60 mrem/hr ( + n) on the top of the overpack OR 600 mrem/hr ( + n) on the side of the overpack excluding inlet and outlet ducts HOSTILE ACTION resulting in loss of physical control of the HOSTILE ACTION within the PROTECTED AREA HOSTILE ACTION within the OWNER CONTROLLED AREA or Confirmed SECURITY CONDITION or threat facility airborne attack threat within 30 minutes HG1.1 1 2 3 4 5 DEF HS1.1 1 2 3 4 5 DEF HA1.1 1 2 3 4 5 DEF HU1.1 1 2 3 4 5 DEF A HOSTILE ACTION is occurring or has occurred within A HOSTILE ACTION is occurring or has occurred within A HOSTILE ACTION is occurring or has occurred within A SECURITY CONDITION that does not involve a the PROTECTED AREA as reported by the Security Shift the PROTECTED AREA as reported by the Security the OWNER CONTROLLED AREA as reported by the HOSTILE ACTION as reported by Security Shift 1 Supervisor AND EITHER of the following has occurred:

Any of the following safety functions cannot be Shift Supervisor Security Shift Supervisor OR A validated notification from NRC of an aircraft attack Supervisor OR Notification of a credible security threat directed at the site Security controlled or maintained threat within 30 min. of the site OR Reactivity A validated notification from the NRC providing information RPV water level of an aircraft threat RCS heat removal OR Damage to spent fuel has occurred or is IMMINENT Seismic event greater than OBE levels 2 None None None HU2.1 1 2 3 4 5 Seismic event greater than Operating Basis Earthquake DEF Seismic Event (OBE) as indicated by peak accelerations > 0.05g vertical or

> 0.08g horizontal on D30-R800 Active Seismic Playback Printer Notes Hazardous event Note 1: The Emergency Director should declare the event promptly upon determining that time limit has HU3.1 1 2 3 4 5 DEF been exceeded, or will likely be exceeded A tornado strike within the PROTECTED AREA Note 2: If an ongoing release is detected and the release start time is unknown, assume that the release HU3.2 1 2 3 4 5 DEF duration has exceeded the specified time limit Internal room or area FLOODING of a magnitude sufficient to require manual or automatic electrical 3 Note 3: If the effluent flow past an effluent monitor is known to have stopped, indicating that the release path is isolated, the effluent monitor None None isolation of a SAFETY SYSTEM component required for the current operating mode Natural or reading is no longer VALID for classification HU3.3 1 2 3 4 5 DEF Tech.

purposes Hazard Movement of personnel within the PROTECTED AREA Note 4: The pre-calculated effluent monitor values is impeded due to an offsite event involving hazardous presented in EALs RA1.1, RS1.1 and RG1.1 materials (e.g., an offsite chemical spill or toxic gas should be used for emergency classification release) assessments until the results from a dose HU3.4 1 2 3 4 5 DEF assessment using actual meteorology are available A hazardous event that results in on-site conditions sufficient to prohibit the plant staff from accessing the Note 5: If the equipment in the listed room or area was site via personal vehicles (Note 7) already inoperable or out-of-service before the event occurred, then no emergency FIRE potentially degrading the level of safety of the plant classification is warranted Note 6: If CONTAINMENT CLOSURE is re-established Table H-1 Fire Areas HU4.1 1 2 3 4 5 DEF prior to exceeding the 30-minute time limit, A FIRE is not extinguished within 15 min. of any of the declaration of a General Emergency is not Reactor Building following FIRE detection indications (Note 1):

required Report from the field (i.e., visual observation)

Auxiliary Building Note 7: This EAL does not apply to routine traffic Receipt of multiple (more than 1) fire alarms or Reactor Building Steam Tunnel indications impediments such as fog, snow, ice, or vehicle H Note 8:

breakdowns or accidents A manual scram action is any operator action, RHR Complex 4160V Duct banks between Field verification of a single fire alarm AND Hazards RHR Complex and Auxiliary The FIRE is located within any Table H-1 area or set of actions, which causes the control rods Building HU4.2 1 2 3 4 5 DEF to be rapidly inserted into the core, and does not include manually driving in control rods or Receipt of a single fire alarm (i.e., no other indications of 4 Note 9:

implementation of boron injection strategies For manual scram actions, the reactor scram None a FIRE)

AND Fire The fire alarm is indicating a FIRE within any Table H-1 pushbuttons, taking the Reactor Mode Switch to area Shutdown or manual initiation of ARI on COP AND H11-P603 are the only methods applicable to The existence of a FIRE is not verified within 30 min. of this EAL alarm receipt (Note 1)

Note 10: Credit may be taken for one of the four CTGs as HU4.3 1 2 3 4 5 DEF an onsite AC power source only if the CTG is A FIRE within the plant PROTECTED AREA not already aligned and capable of powering an extinguished within 60 min. of the initial report, alarm or essential bus within 15 min. indication (Note 1)

HU4.4 1 2 3 4 5 DEF A FIRE within the plant PROTECTED AREA that requires firefighting support by an offsite fire response agency to extinguish Gaseous release impeding access to equipment necessary for normal plant operations, cooldown or shutdown Table H-2 Safe Shutdown Rooms/Areas 5 None None HA5.1 1 2 3 4 5 Release of a toxic, corrosive, asphyxiate or flammable gas DEF Room/Area Mode Applicability Hazardous Gases into any Table H-2 rooms or areas Control Room All AND Relay Room All Entry into the room or area is prohibited or impeded (Note 5)

AB3-DC MCC Area 3 Inability to control a key safety function from outside the Control Control Room evacuation resulting in transfer of plant control to RB1-F17 3 Room alternate locations RB1-F11 3 HS6.1 1 2 3 4 5 HA6.1 1 2 3 4 5 DEF 6 None An event has resulted in plant control being transferred from the Control Room to the Dedicated or Remote Shutdown Panels An event has resulted in plant control being transferred from the Control Room to the Dedicated or Remote Shutdown Panels None Control AND Room Control of any of the following key safety functions is not Evacuation reestablished within 15 min. (Note 1):

Reactivity (Mode 1 and 2 only)

RPV water level RCS heat removal Other conditions existing that in the judgment of the Emergency Other conditions existing that in the judgment of the Emergency Other conditions existing that in the judgment of the Emergency Other conditions existing that in the judgment of the Emergency Director warrant declaration of General Emergency Director warrant declaration of Site Area Emergency Director warrant declaration of an Alert Director warrant declaration of a UE HG7.1 1 2 3 4 5 DEF HS7.1 1 2 3 4 5 DEF HA7.1 1 2 3 4 5 DEF HU7.1 1 2 3 4 5 DEF Other conditions exist which in the judgment of the Other conditions exist which in the judgment of the Other conditions exist which, in the judgment of the Other conditions exist which in the judgment of the Emergency Director indicate that events are in progress or Emergency Director indicate that events are in progress or Emergency Director, indicate that events are in progress or Emergency Director indicate that events are in progress or have occurred which involve actual or likely major failures of have occurred which involve an actual or potential have occurred which indicate a potential degradation of the 7 have occurred which involve actual or IMMINENT substantial core degradation or melting with potential for loss of containment integrity or HOSTILE ACTION that plant functions needed for protection of the public or HOSTILE ACTION that results in intentional damage or substantial degradation of the level of safety of the plant or a security event that involves probable life threatening risk level of safety of the plant or indicate a security threat to facility protection has been initiated. No releases of Judgment results in an actual loss of physical control of the facility. malicious acts, (1) toward site personnel or equipment that to site personnel or damage to site equipment because of radioactive material requiring offsite response or monitoring Releases can be reasonably expected to exceed EPA could lead to the likely failure of or, (2) that prevent effective HOSTILE ACTION. Any releases are expected to be limited are expected unless further degradation of SAFETY Protective Action Guideline exposure levels offsite for more access to equipment needed for the protection of the public. to small fractions of the EPA Protective Action Guideline SYSTEMS occurs than the immediate site area. Any releases are not expected to result in exposure levels exposure levels.

which exceed EPA Protective Action Guideline exposure levels beyond the SITE BOUNDARY.

EAL Classification Matrix Page 1 of 3 Modes: 1 2 3 4 5 DEF Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled ALL CONDITIONS Prepared for DTE Energy by: Operations Support Services, Inc. - 9/17/15

EP-101 Enclosure A, Page 2 of 3 101819A EMERGENCY ACTION LEVEL (EAL) CLASSIFICATION MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Prolonged loss of all offsite and all onsite AC power to essential Loss of all offsite and all onsite AC power to essential buses Loss of all but one AC power source to essential buses Loss of all offsite AC power capability to essential buses buses OR loss of all essential AC and vital DC power sources for for 15 minutes or longer for 15 minutes or longer for 15 minutes or longer 15 minutes or longer SG1.1 1 2 3 SS1.1 1 2 3 SA1.1 1 2 3 SU1.1 1 2 3 Loss of all offsite and all onsite AC power capability Loss of all offsite and all onsite AC power capability AC power capability to 4160V essential Division I (64B/64C) Loss of all offsite AC power capability (Table S-1) to 4160V (Table S-1) to 4160V essential Division I (64B/64C) and (Table S-1) to 4160V essential Division I (64B/64C) and and Division II (65E/65F) reduced to a single power source essential Division I (64B/64C) and Division II (65E/65F) for Division II (65E/65F) (Note 10) Division II (65E/65F) for 15 min. (Note 1, 10) (Table S-1) for 15 min. (Note 1, 10) 15 min. (Note 1, 10)

AND EITHER of the following:

1 Restoration of at least one essential division within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is not likely (Note 1) Table S-1 AC Power Sources Loss of RPV water level CANNOT be restored and Essential maintained > -25 in. Offsite:

AC Power System Service Transformer 64 (Div I)

SG1.2 1 2 3 System Service Transformer 65 (Div II)

Loss of all offsite and all onsite AC power capability Onsite:

(Table S-1) to 4160V essential Division I (64B/64C) and EDGs 11/12 (Div I)

Division II (65E/65F) for 15 min. (Note 1, 10) EDGs 13/14 (Div II)

AND CTG 11-1, 11-2, 11-3 or 11-4 Degraded voltage (< 105 VDC) on both 130 VDC system vital buses for 15 min. (Note 1)

Loss of all vital DC power for 15 minutes or longer 2 None SS2.1 1 2 3 None None Loss of Degraded voltage (< 105 VDC) on both 130 VDC system Vital DC vital buses for 15 min. (Note 1)

Power UNPLANNED loss of Control Room indications for 15 minutes or UNPLANNED loss of Control Room indications for 15 minutes or longer with a significant transient in progress longer 3 None SA3.1 1 2 3 SU3.1 1 2 3 Loss of None An UNPLANNED event results in the inability to monitor one An UNPLANNED event results in the inability to monitor Control or more Table S-2 parameters from within the Control Room one or more Table S-2 parameters from within the Room for 15 min. (Note 1) Control Room for 15 min. (Note 1)

Indications AND Any Table S-3 transient event in progress Table S-2 Safety System Parameters Reactor coolant activity greater than Technical Specification allowable limits Reactor power SU4.1 1 2 3 4 None RPV water level RPV pressure None Offgas radiation monitor D11-K601A or D11-K601B high-high alarm (ARP 3D12) (Note 11)

RCS Table S-4 Communication Methods Primary Containment pressure 3 SU4.2 1 2 Activity Torus water level Sample analysis indicates that a reactor coolant activity value System Onsite Offsite NRC Torus temperature is > an allowable limit specified in Technical Specifications (Note 12)

Administrative Telephones X X X Table S-3 Transient Events RCS leakage for 15 minutes or longer RERP Emergency Telephones X X X Satellite Phones X X SU5.1 1 2 3 S Federal Telephone System (ENS) X X Automatic or manual runback > 25%

RCS unidentified or pressure boundary leakage > 10 gpm System Malfuncti 5 None Automatic Ring Lines MI State Radios (800 MHz)

None X

X thermal reactor power Electrical load rejection electrical load None > 25% full for 15 min.

OR RCS RCS identified leakage > 25 gpm for 15 min.

on. Leakage Plant Radio System X Reactor scram OR Hi-Com (PA System) X ECCS actuation Leakage from the RCS to a location outside Primary Containment > 25 gpm for 15 min.

Thermal power oscillations > 10%

(Note 1)

Inability to shut down the reactor causing a challenge to RPV Automatic or manual scram fails to shut down the reactor, and Automatic or manual scram fails to shut down the reactor water level or RCS heat removal subsequent manual actions taken at the reactor control consoles are not successful in shutting down the reactor SS6.1 1 2 SA6.1 1 2 SU6.1 1 2 An automatic or manual scram fails to shut down the reactor An automatic or manual scram fails to shut down the reactor An automatic scram did not shut down the reactor after any AND AND RPS setpoint is exceeded All actions to shut down the reactor are not successful as AND 6 None indicated by reactor power > 3%

AND EITHER of the following conditions exist:

Manual scram actions taken at COP H11-P603 are not successful in shutting down the reactor as indicated by reactor power > 3% (Note 8, 9)

A subsequent automatic scram or manual scram action taken at COP H11-P603 is successful in shutting down the RPS reactor as indicated by reactor power < 3% (Note 8, 9)

Failure RPV water level cannot be restored and maintained SU6.2 Notes > -25 in.

Torus water temperature and RPV pressure cannot 1 2 A manual scram did not shutdown the reactor after any be maintained below the Heat Capacity Limit (HCL) manual scram action was initiated Note 1: The Emergency Director should declare the event promptly upon determining that time limit has AND been exceeded, or will likely be exceeded A subsequent automatic scram or manual scram action Note 8: A manual scram action is any operator action, taken at COP H11-P603 is successful in shutting down or set of actions, which causes the control rods the reactor as indicated by reactor power < 3% (Note 8, 9) to be rapidly inserted into the core, and does not include manually driving in control rods or Loss of all onsite or offsite communications capabilities implementation of boron injection strategies SU7.1 1 2 3 7 Note 9: For manual scram actions, the reactor scram pushbuttons, taking the Reactor Mode Switch to Shutdown or manual initiation of ARI on COP None None Loss of all Table S-4 onsite communication methods OR Loss of H11-P603 are the only methods applicable to Comm. Loss of all Table S-4 offsite communication methods this EAL OR Note 10: Credit may be taken for one of the four CTGs as Loss of all Table S-4 NRC communication methods an onsite AC power source only if the CTG is None Table S-5 Hazardous Events already aligned and capable of powering an Hazardous event affecting a SAFETY SYSTEM required for the essential bus within 15 min. current operating mode Seismic event (earthquake)

Note 11: Consistent with Technical Specification 3.7.5, Internal or external FLOODING event 8 this EAL is applicable at all times while in Mode 1, Mode 2 or in Mode 3 with any main steam line not isolated and steam jet air ejector in High winds or tornado strike FIRE SA8.1 1 2 3 The occurrence of any Table S-5 hazardous event AND Hazardous operation EXPLOSION EITHER of the following: None Event Event damage has caused indications of degraded Affecting Note 12: Consistent with Technical Specification 3.4.7, Other events with similar hazard Safety performance in at least one train of a SAFETY this EAL is applicable at all times while in Mode characteristics as determined by the Systems SYSTEM required for the current operating mode 1, Mode 2 or in Mode 3 with any main steam Shift Manager The event has caused VISIBLE DAMAGE to a line not isolated SAFETY SYSTEM component or structure required for the current operating mode 3

F FG1.1 1 Loss of any two barriers 2 FS1.1 1 2 3 Loss or potential loss of any two barriers (Table F-1)

FA1.1 1 2 3 Any loss or any potential loss of either Fuel Clad or RCS None Fission Product AND barrier (Table F-1)

Barrier Degradation Loss or potential loss of the third barrier (Table F-1)

Table F-1 Fission Product Barrier Threshold Matrix Fuel Clad Barrier Reactor Coolant System Barrier Primary Containment Barrier Loss Potential Loss Loss Potential Loss Loss Potential Loss A. Inadequate core cooling as A. RPV water level cannot be restored A. RPV water level cannot be restored A. Inadequate core cooling as

1. RPV Water indicated by any of the following: and maintained above 0 in. (TAF) or and maintained above 0 in. (TAF) or indicated by any of the following:

Level

1. RPV water level cannot be cannot be determined. cannot be determined. 1. RPV water level cannot be restored and maintained > -48 restored and maintained > -48 in. with 5725 gpm Core Spray in. with 5725 gpm Core Spray loop flow loop flow OR OR None None 2. RPV water level cannot be

2. RPV water level cannot be restored and maintained > -25 in restored and maintained > -25 in with < 5725 gpm Core Spray with < 5725 gpm Core Spray loop flow loop flow OR OR

3. RPV water level cannot be 3. RPV water level cannot be determined and core damage is determined and core damage is occurring occurring

2. RCS Leak Rate A. UNISOLABLE break in any of the A. UNISOLABLE primary system leakage A. UNISOLABLE primary system leakage following: into Secondary Containment that into Secondary Containment that Main Steam Line results in exceeding EITHER of the results in exceeding EITHER of the HPCI Steam Line following: following:

RCIC Steam Line 1. One or more Secondary 1. One or more Secondary RWCU Containment Control Max Normal Containment Control Max Safe None None Feedwater Operating Temperatures (EOP Operating Temperatures (EOP None Table 12) Table 12)

OR OR OR

2. One or more Secondary 2. Exceeding Secondary B. Emergency RPV Depressurization Containment Control Max Normal Containment Control Max Safe is required Operating Area Radiation Levels Operating Area Radiation Level on (EOP Table 14) channel 14 ARM (RBSB Torus Room)

A. Drywell pressure > 1.68 psig due to A. UNPLANNED rapid drop in Primary A. Primary Containment Pressure

3. PC Conditions RCS leakage Containment pressure following > 62 psig Primary Containment pressure rise OR OR B. > 6% H2 AND > 5% O2 in EITHER the drywell or None None None B. Primary Containment pressure suppression chamber response not consistent with LOCA conditions OR C. EOP Heat Capacity Limit (HCL) exceeded

4. PC Radiation / A. CHRRM reading > 2.25E+3 R/hr A. CHRRM reading > 8.72E+1 R/hr A. CHRRM reading > 1.79E+4 R/hr RCS Activity OR None None None B. Primary coolant activity > 300

µCi/gm DEI-131 A. UNISOLABLE direct downstream

5. PC Integrity or pathway to the environment exists Bypass after Primary Containment isolation signal None None None None OR None B. Intentional Primary Containment venting, irrespective of offsite radioactivity release rates, per EOPs

6. Emergency A. Any condition in the opinion of the A. Any condition in the opinion of the A. Any condition in the opinion of the A. Any condition in the opinion of the A. Any condition in the opinion of the A. Any condition in the opinion of the Director Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates Emergency Director that indicates loss Emergency Director that indicates Judgment loss of the fuel clad barrier potential loss of the Fuel Clad barrier loss of the RCS barrier potential loss of the RCS barrier of the Primary Containment barrier potential loss of the Primary Containment barrier EAL Classification Matrix Page 2 of 3 Modes: 1 2 3 4 5 DEF HOT CONDITIONS Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled (RCS > 200°F)

Prepared for DTE Energy by: Operations Support Services, Inc. - 9/17/15

EP-101 Enclosure A, Page 3 of 3 101819A EMERGENCY ACTION LEVEL (EAL) CLASSIFICATION MATRIX GENERAL EMERGENCY SITE AREA EMERGENCY ALERT UNUSUAL EVENT Loss of RPV inventory affecting fuel clad integrity with Loss of RPV inventory affecting core decay heat removal Loss of RPV inventory Unplanned loss of RPV inventory for 15 minutes or longer Containment challenged capability CG1.1 4 5 CS1.1 4 5 CA1.1 4 5 CU1.1 4 5 RPV water level < 0 in. for 30 min. (Note 1) CONTAINMENT CLOSURE not established Loss of RPV inventory as indicated by RPV water level UNPLANNED loss of reactor coolant results in RPV water AND AND < 111 in. above TAF (Level 2) level below the established control band for 15 min.

Any of the following indications of containment challenge: RPV water level < 32 in. above TAF (Level 1) (Note 1)

CONTAINMENT CLOSURE not established (Note 6)

Primary Containment hydrogen concentration > 6% CS1.2 4 5 CA1.2 4 5 CU1.2 4 5 UNPLANNED increase in Primary Containment pressure CONTAINMENT CLOSURE established RPV water level cannot be monitored for 15 min. (Note 1) RPV water level cannot be monitored Exceeding Secondary Containment Control Max Safe AND AND AND 1 Operating Area Radiation Level on channel 14 ARM (RBSB Torus Room)

RPV water level < 0 in.

CS1.3 4 5 UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory UNPLANNED increase in any Table C-1 sump or tank levels due to a loss of RPV inventory RPV Level CG1.2 4 5 RPV water level cannot be monitored for 30 min. (Note 1)

RPV water level cannot be monitored for 30 min. (Note 1) AND Table C-1 Sumps & Tanks AND Core uncovery is indicated by EITHER of the following:

Core uncovery is indicated by EITHER of the following: RB5 Refuel Floor Hi Range ARM (Ch. 18) alarm RB5 Refuel Floor Hi Range ARM (Ch. 18) alarm Drywell Floor Drain Sump UNPLANNED increase in any Table C-1 sump or UNPLANNED increase in any Table C-1 sump or tank tank levels due to a loss of RPV inventory Drywell Equipment Drain Sump levels due to a loss of RPV inventory RB Floor Drain Sumps AND RB Equipment Drain Sumps Any of the following indications of containment challenge:

Torus CONTAINMENT CLOSURE not established (Note 6)

Visual Observation Primary Containment hydrogen concentration > 6%

UNPLANNED increase in Primary Containment pressure Exceeding Secondary Containment Control Max Safe Operating Area Radiation Level on channel 14 ARM (RBSB Torus Room)

Loss of all offsite power and all onsite AC power to essential Loss of all but one AC power source to essential buses for 15 Table C-2 AC Power Sources buses for 15 minutes or longer minutes or longer 2 None Offsite:

System Service Transformer 64 (Div I)

System Service Transformer 65 (Div II)

CA2.1 4 5 DEF Loss of all offsite and all onsite AC power capability (Table CU2.1 4 5 DEF AC power capability to 4160V essential Division I (64B/64C)

Loss of C-2) to 4160V essential Division I (64B/64C) and Division II and Division II (65E/65F) reduced to a single power source Essential Onsite: (65E/65F) for 15 min. (Note 1, 10) (Table C-2) for 15 min. (Note 1, 10)

AC Power EDGs 11/12 (Div I)

EDGs 13/14 (Div II)

C CTG 11-1, 11-2, 11-3 or 11-4 Cold SD/

Refueling Table C-3 RCS Reheat Duration Thresholds Inability to maintain plant in cold shutdown UNPLANNED increase in RCS temperature System Malfunct.

If an RCS heat removal system is in operation within this CA3.1 4 5 CU3.1 4 5 3 None time frame and RCS temperature is being reduced the EAL is not applicable UNPLANNED increase in RCS temperature to > 200°F for

> Table C-3 duration (Note 1)

UNPLANNED increase in RCS temperature to > 200°F RCS Containment Heat-up RCS Status Temp. Closure Status Duration CA3.2 4 5 CU3.2 4 5 Intact N/A 60 min.

UNPLANNED RPV pressure increase > 10 psig Loss of all RCS temperature and RPV level indication for 15 min. (Note 1) established 20 min.

Not intact Loss of Vital DC power for 15 minutes or longer 4 None not established None 0 min.

Table C-4 Communication Methods CU4.1 4 5 Loss of Degraded voltage (< 105 VDC) on required 130 VDC Vital DC system vital buses for 15 min. (Note 1)

Power System Onsite Offsite NRC Administrative Telephones X X X Loss of all onsite or offsite communications capabilities RERP Emergency Telephones X X X CU5.1 4 5 DEF Satellite Phones X X 5 None None Federal Telephone System (ENS)

Automatic Ring Lines X

X X Loss of all Table C-4 onsite communication methods OR Loss of Comm. MI State Radios (800 MHz) X Loss of all Table C-4 offsite communication methods Plant Radio System X OR Hi-Com (PA System) X Loss of all Table C-4 NRC communication methods Hazardous event affecting a SAFETY SYSTEM required for the Table C-5 Hazardous Events current operating mode Seismic event (earthquake) CA6.1 4 5 6 None Internal or external FLOODING event High winds or tornado strike The occurrence of any Table C-5 hazardous event AND None Hazardous EITHER of the following:

Events FIRE Event damage has caused indications of degraded Affecting EXPLOSION performance in at least one train of a SAFETY Safety Other events with similar hazard SYSTEM required for the current operating mode Systems characteristics as determined by the The event has caused VISIBLE DAMAGE to a Shift Manager SAFETY SYSTEM component or structure required for the current operating mode Notes Note 1: The Emergency Director should declare the event promptly upon determining that time limit has been exceeded, or will likely be exceeded Note 6: If CONTAINMENT CLOSURE is re-established prior to exceeding the 30-minute time limit, declaration of a General Emergency is not required.

Note 10: Credit may be taken for one of the four CTGs as an onsite AC power source only if the CTG is already aligned and capable of powering an essential bus within 15 min.

EAL Classification Matrix Page 3 of 3 Modes: 1 2 3 4 5 DEF COLD CONDITIONS Power Operations Startup Hot Shutdown Cold Shutdown Refueling Defueled (RCS 200°F)

Prepared for DTE Energy by: Operations Support Services, Inc. - 9/17/15

ML22158A215

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ML22158A215

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