ML16280A464

2016-09 Draft Written Exam

ML16280A464
Person / Time
Site:	River Bend
Issue date:	09/30/2016
From:	Vincent Gaddy Operations Branch IV
To:	Entergy Operations
References
Download: ML16280A464 (219)

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Examination Outline Cross-Reference Level RO 271000 Offgas Tier # 2 Group # 2 Ability to monitor automatic operations of the K/A # A3.07 OFFGAS SYSTEM including: Process Rating 3.4 radiation monitoring system indications Question 1 The major source of radiation monitored by the OFF-GAS radiation monitors is primarily

________. When the OFF-GAS POST TRT HIGH RADIATION alarm is received the OFF-GAS system will automatically ____________.

A. gamma radiation from O-19 and N-16 place the charcoal beds on line for removal of radioactive particulates from the off-gas B. gamma radiation from O-19 and N-16 shut OFF-GAS DISCH TO VENT VLV, 1N64-F060, to isolate the off-gas stream from discharging to the stack exhaust ventilation C. gamma radiation emitted by radioactive particles place the charcoal beds on line for removal of radioactive particulates from the off-gas D. gamma radiation emitted by radioactive particles shut OFF-GAS DISCH TO VENT VLV, 1N64-F060, to isolate the off-gas stream from discharging to the stack exhaust ventilation Answer: C Explanation:

A. Is wrong because the sample is extracted from the line at a point where the shorter lived radioactive nuclei, principally O-19 and N-16, have had sufficient time to decay, and contribute only minimally to the gross radioactivity.

The charcoal beds would automatically be placed in service on receipt of this alarm.

B. Is wrong because the sample is extracted from the line at a point where the shorter lived radioactive nuclei, principally O-19 and N-16, have had sufficient time to decay, and contribute only minimally to the gross radioactivity The OFF-GAS DISCH TO VENT VLV, 1N64-F060 is isolated if An isolation of the Off-gas System will occur if both channels receive HI-HI-HI and/or INOP signals from the trip units C. Correct

D. Is wrong because OFF-GAS DISCH TO VENT VLV, 1N64-F060 is isolated if An isolation of the Off-gas System will occur if both channels receive HI-HI-HI and/or INOP signals from the trip units Technical

References:

System Training Manual, R-STM-0511 Rev 15, Radiation Monitoring Systems References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0511 Obj X Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b) 7

Examination Outline Cross-Reference Level RO 295032 High Secondary Containment Area Tier # 1 Temperature Group # 2 K/A # EK3.03 Knowledge of the reasons for the following Rating 3.8 responses as they apply to high secondary containment area temperature: Isolating affected systems Question 2 You are performing RCIC surveillance when an alarm indicating a room high temperature is received.

Temperature readings are as follows:

RCIC Equipment Room Ambient Temperature is 184 degrees rising RHR/RCIC Area Temperature is 124 degrees and rising RHR Equipment Room Ambient Temperature is 110 and rising RHR area Temperature is 108 degrees and rising Based on this indications provided what isolation signals have been generated?

A. Group 2 valves only B. Groups 2 and 3 valves only C. Groups 5, 14 and 17 only D. Groups 2, 5, 14 and 17 only Answer: A Explanation:

A. Correct B. Incorrect, because high RCIC equipment room temp of 182 degrees will result in group 2 valves isolating only, the distractor is plausible because the high temperature in the RCIC room would give a group 2 valve isolation and if RCIC steam supply pressure was low < 60 psig concurrent with a high drywell pressure (1.68) you would receive both group 2 and 3 valve isolations.

C. Incorrect, because only the RCIC room ambient temperature has exceeded the actuation set point of 182 degrees which will isolate group 2 valves only. The distractor is plausible because if RHR Area Temperature was > 130.9 degrees you would get Groups 5,14 and 17 valve isolations D. Incorrect, because only the RCIC room ambient temperature has exceeded the actuation set point for Group 2 isolations if the RHR/RCIC Area high temperature had exceeded 130.9 or the RHR Equipment Room Ambient temperature exceeded 117 degrees groups 2, 5, 14 anbd17 valves would have received an isolation signal.

Technical

References:

AOP-0003 Rev 34 page References to be provided to applicants during exam: None.

Learning Objective: RLP-HLO-0522, AOP 0003, Automatic Isolations, Objective C Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b) 7

4 Examination Outline Cross-Reference Level RO 218000 ADS Tier # 2 Group # 1 Knowledge of ADS design features and/or K/A # K4.03 interlocks which provide for the following: Rating 3.8 Logic control Question 3 The following conditions exist:

• A leak inside the drywell has occurred.

• All RHR pumps are running.

• Automatic Depressurization System automatically actuated at 105 seconds and all

• ADS SRVs are open.

• RPV water level is now steady at -150 inches.

• Drywell pressure peaked at 1.5 psid and is now lowering.

• RPV pressure is 200 psig.

If both Div 1 and Div 2 ADS TIMER/LEVEL 3 SEAL IN RESET buttons are depressed and then released, which of the following describes the result on the Automatic Depressurization System?

The ADS SRVs will . . .

A. close and then reopen after 5 minutes plus 105 seconds.

B. close and then reopen after 105 seconds.

C. close and remain closed.

D. remain open.

Answer: A Explanation:

A. Correct- they will close when the reset pushbuttons (ADS TIMER/LEVEL 3 SEAL-IN RESET pushbuttons, S13A(B)) are operated, which resets the 105-second timers and because the low water level signal is still in (-143 inches) the 5 minute timer also starts. Unless level goes above -143 inches they will reopen once the timer is out.

B. Incorrect-see A - credible because it you dont understand the logic or forget about the 5 minute timer you might pick this distracter.

C. Incorrect- if you dont recognize the initiation signal is still active for RPV level then you might think this is correct.

D. Incorrect - if you dont understand the logic of the reset seal-in buttons you might think that you have to use the manual buttons on the SRVs to close then once the reset is selected, similar to reset on ECCS equipment, but this is not true for this logic circuit.

Technical

References:

STM-0202 Rev 2, pages 7 and 13-15, and Figures 2 and 4.

References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0202 Obj H3 and L2 Question Source: Bank # NRC 2003 (Q14)

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New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental H3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b) 8

4 Examination Outline Cross-Reference Level RO 295003 Partial or Complete Loss of AC Tier # 1 Group # 1 Ability to operate and/or monitor the following K/A # AA1.04 as they apply to partial or complete loss of Rating 3.6 A.C. power: D.C. electrical distribution system Question 4 The reactor is operating at 100% power.

A station blackout occurs.

AC power restoration is expected to take 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

The Station Blackout Diesel Generator should be aligned to 125 VDC bus _____(1)_____,

via BYS-CHGR1D, 125 VDC BACKUP BATT CHGR, in accordance with AOP-50, Station Blackout.

Following AC power restoration, the normal charger for this bus can recharge the associated battery from full discharge to full charge within a maximum of ___(2)___ hours while simultaneously handling steady state DC loads.

A. (1) ENB-SWG01A (Div I)

(2) 8 B. (1) ENB-SWG01A (Div I)\

(2) 24 C. (1) E22-PNLS001 (DIV III)

(2) 8 D. (1) E22-PNLS001 (DIV III)

(2) 24 Answer: B Explanation:

A is wrong because 125 VDC bus ENB-SWG01A (Div I) battery ENB-BAT01A can be recharged within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, not 8.

B is correct. See explanation below.

C is wrong because AOP-50 directs the backup battery charger to be aligned to Div I ENB-SWG01A, not Div III E22-PNLS001. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is the correct recharge time for E22-PNLS001, but this is not the correct bus to be aligned to.

D is wrong because AOP-50 directs the backup battery charger to be aligned to Div I ENB-SWG01A, not Div III E22-PNLS001. 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> is the correct recharge time for E22-PNLS001, not 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Technical

References:

AOP-50 Rev 55, Station Blackout, Revision 55.

R-STM-0305 Rev 7, DC Distribution.

References to be provided to applicants during exam: None.

Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b) 8

4 Examination Outline Cross-Reference Level RO 295026 Supp. Pool High Water Temp Tier # 1 Group # 1 Knowledge of the interrelations between K/A # EK2.05 suppression pool high water temperature and Rating 3.0 the following: Containment pressure: Mark-III Question 5 During an accident, with RPV pressure at 800 psig, the Average Suppression Pool Temperature reaches the Heat Capacity Temperature Limit (HCTL) and is still increasing. At this point the EOPs direct operators to Emergency Depressurize. An ED at this point will prevent exceeding CERTAIN CONTAINMENT LIMITS. Which of the following is the MOST LIMITING containment parameter for these conditions?

A. The Primary Containment Pressure Limit.

B. The Pressure Suppression Pressure Limit.

C. The Drywell Design Temperature Limit.

D. The Primary Containment Temperature Limit.

Answer: B Explanation:

A. Incorrect. The primary containment pressure limit is around 30 psig and the PSP limit is 3.2 to 5 psig and would be reached first for these conditions B. Correct - the PSP limit (variable between 3.2 to 5.5 psig SAFE region) is reached before any of the other variables and is the main concern for these conditions to be able to reject the energy from the RPV during an ED.

C. Incorrect DW Temperature is a concern but not the first concern reached for these conditions.

D. Incorrect, Containment temp is a concern but not the major concern for these conditions.

Technical

References:

EOP-2, Primary Containment Control, Revision 16 EOP bases document, Step CP-5, page B-8-14.

References to be provided to applicants during exam: None.

Learning Objective: HLO-0514 Obj 5 Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2

Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b) 9

Examination Outline Cross-Reference Level RO 700000 Generator Voltage and Electric Grid Tier # 1 Disturbances Group # 1 K/A # AA1.02 Ability to operate and/or monitor the following Rating 3.8 as they apply to generator voltage and electric grid disturbances: Turbine/generator controls Question 6 The Reactor is operating at 100% power The Main Generator Voltage Regulator is in Manual, due to a failure associated with the AC regulator.

Alarm P808-86A-H01 GRID DISTRIBUNCE is in alarm THE SOC , Pine Bluff, Reports that the Grid is stable at 223 KV and cannot be maintained above 224.25 KV Per AOP-0064, Degraded Grid, you are directed to trend ____________ and A. The PDS data for Main Generator MVARs (PDS Point - SPGEA01)

Perform a Normal Start of all Three Emergency Diesel Generators and then parallel with offsite power and disconnect the bus from the grid.

B. The PDS data for Main Generator Voltage output (PDS Point SPGEA02)

Perform a Normal Start of all Three Emergency Diesel Generators and then parallel with offsite power and disconnect the bus from the grid.

C. The PDS data for Main Generator MVARs (PDS Point - SPGEA01)

Perform an Emergency Start of all three Emergency Diesel Generators and then open the supply breakers to each divisional safety related busses.

D. The PDS data for Main Generator Voltage output (PDS Point SPGEA02)

Perform an Emergency Start of all three Emergency Diesel Generators and then open the supply breakers to each divisional safety related busses Answer: B Explanation:

A. Incorrect, Monitoring Main Generator MVARS is correct if the Main Generator Voltage regulator is in automatic.

B. Correct C. Incorrect, Monitoring the Main Generator MVARs is correct if voltage Regulator is in automatic, and an emergency start would be correct if the grid voltage was not stable D. Incorrect, Emergency Start of the diesels would be correct if the grid voltage was not stable E.

Technical

References:

AOP-0064, Degraded Grid References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b) 10 1.1 IF it is determined that grid voltage can not be maintained above 224.25 KV (97.5%), THEN perform the following:

1.1.1. Consult with SOC/TOC regarding stability of the grid near RBS.

1.1.2. IF SOC/TOC indicates that the Fancy Point is sufficiently stable, THEN [5.5.3]

NOTE It is preferred that Div 3 DG is transferred first and the division that has HVK running transferred last to prevent alternating HVK twice.

1. Perform a Normal start of all three Emergency Diesel Generators.

NOTE If the synchroscope for the Diesel Generator(s) shows erratic movement, then the grid should be considered unstable and synchronizing the Diesel Generator to the grid should be avoided.

2. Use the synchroscopes used for paralleling the Diesel Generators to the safety busses as an instantaneous indication of instability on the grid.

NOTE An unstable grid will result in swings in at least one of the following data points.

3. Use the following to trend the grid stability before paralleling the Diesel Generator to the grid by comparing the trends with historical data, inspecting for erratic behavior such as sudden spikes and dips:

• IF the Main Generator Voltage Regulator is in Auto, THEN trend the PDS data for Main Generator MVARs (PDS Point -

SPGEA02).

• IF the Main Generator Voltage Regulator is in Manual, THEN trend the PDS data for Main Generator Voltage output (PDS Point - SPGEA01).

• Trend PDS data for the Main Turbine Speed (PDS Point -

N31EA001).

4. IF the grid is verified stable, THEN Parallel the Emergency Diesel Generators with offsite power AND disconnect the bus from the grid.

1.1.3. IF SOC/TOC OR plant indications used in step 5.5.2 indicate that the Fancy Point is NOT sufficiently stable, THEN:

1. Emergency Start each diesel generator.

2. Verify diesel system parameters are normal.

3. Ensure all isolations are reset for the applicable safety related bus prior to continuing to the next bus. [5.5.3.6]

CAUTION IF RPS is being supplied from the Alternate source, THEN deenergizing the divisional safety related bus will cause a RPS half SCRAM and Inboard or Outboard BOP isolation.

NOTE It is preferred that Div 3 DG is transferred first and the division that has HVK running transferred last to prevent alternating HVK twice.

4. Open supply breaker to a single divisional safety related bus.

5. Restore isolations and systems as necessary.

• Verify alarms caused by loss of power clear upon power restoration.

• Restore RWCU and SPC, IF isolated on loss of power to leak detection.

• Reset half isolations for RWCU and MSL Drains per AOP-0003, Automatic Isolations.

• Restart Rad Monitors.

• Reset RPS Alternate EPA Breakers.

• Alternate Divisions of HVK if necessary.

• Reset SRV Accoustic Monitors (Division 2).

• Reset Bearing Lift Pumps following loss of power per ARP-680-15A-A2 (Division 2).

• Restore Generator Seal Oil to normal lineup per SOP-0019, Generator Seal Oil System (Division 2).

6. Repeat step 5.5.3.3 through 5.5.3.6 for remaining safety related busses. [5.5.3.6]

Examination Outline Cross-Reference Level RO Tier # 3 Group #

Ability to identify and interpret diverse K/A # 2.1.45 indications to validate the response of another Rating 4.1 indication Question 7 As the ATC you want to verify that an RPV level instrument is working correctly. Because each instrument for RPV level may display a slightly different value, you use the sub-system in ERIS called _____ to validate the parameters value.

A. TAMARIS B. POWERPLEX C. RTAD D. PPC Answer: C Explanation:

A. Incorrect, TAMARIS does not perform parameter validation, RTAD within ERIS does this.

B. Incorrect, POWERPLEX does not perform parameter validation, RTAD within ERIS does this.

C. Correct, per reference, RTAD is the sub system within ERIS that performs parameter validation D. Incorrect, PPC does not perform parameter validation.

Technical

References:

STM-0514, rev 11, page 12.

References to be provided to applicants during exam: None.

Learning Objective:

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2

Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO 215004 Source Range Monitor Tier # 2 Group # 1 Ability to manually operate and/or monitor in K/A # A4.07 the control room: A4.07 Verification of proper Rating 3.4 functioning/ operability Question 8 The reactor has been shut down for an extended refueling outage.

SRM indications are as follows:

Consider getting a picture with the SRM A reads 2X10 cps 1 values shown instead of a table SRM B reads 4X101 cps SRM C reads 5X101 cps SRM D reads 2X105 cps A rod withdraw block is currently in due to which of the following?

A. Source Range Downscale Only B. Source Range Monitor High Flux Only C. Source Range Monitor High Flux and SRM Inoperable Only D. Low Source Range Counts and Source Range Monitor High Flux Answer: D Explanation:

A. Incorrect because Channel SRM channel greater than 1X105 cps will also cause a read withdrawal block B. Incorrect because low source range monitor of < 3 cps will also generate a rod withdrawal block C. Incorrect, because low source range monitor of < 3cps will also generate a rod withdrawal block D. Correct Technical

References:

R-STM-0503, Rev 9 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content: 55.41(b) 2 From R-STM-0503 SRM Downscale This condition indicates possible channel malfunction or insufficient source neutron level for a safe reactor startup. It generates a control rod withdrawal block when SRM counts fall below 3.0 cps. This control rod withdrawal block is bypassed when any of the following conditions exist:

Associated IRMs on Range 3 or above Reactor Mode Switch in RUN Associated SRM channel bypassed SRM High Flux This condition indicates a neutron flux level in excess of that allowed for a specific plant operating condition. It generates a control rod withdrawal block when SRM counts exceeds 1 x 105 cps. The control rod withdrawal block is bypassed when any of the following conditions exist:

• Associated IRMs on Range 8 or above

• Reactor Mode Switch in RUN

• Associated SRM channel bypassed SRM Inoperative The Inoperative trip signal causes the same actions as the High Flux trip. This condition prevents control rod withdrawal under the following circumstances:

• SRM module is unplugged

• High voltage power supply output less than 95%

• SRM Mode Switch not in operate The rod withdrawal block is bypassed if any of the following conditions exist:

• Associated IRMs on Range 8 or above

• Reactor Mode Switch in RUN

Examination Outline Cross-Reference Level RO 295005 Main Turbine Generator Trip Tier # 1 Group # 1 2.1.19 Ability to use plant computers to K/A # 2.1.19 evaluate system or component status. Rating 3.9 Question 9 The plant is operating at 100% power when a turbine trip occurs. After 15 seconds from the time of the turbine trip, the following data is observed on the boards and Plant Process Computer (PPC):

• Condenser vacuum reads 26.5

• The Exciter Field Breaker is closed

• Main Generator Gross MWe indicates 0 MWe According to AOP-0002, MAIN TURBINE AND GENERATOR TRIPS, the main generator output breakers should be _______ and the specific limit for motoring the main generator is _______.

A. open; 90 seconds B. closed; 90 seconds C. closed; 5 minutes D. open; 5 minutes Answer: D. open; 5 minutes Explanation:

A is wrong because 90 seconds is incorrect - see B.

B is wrong because the anti-motoring relay actuates with a 5 sec time delay; plausible if applicants only consider the second relay which has a 30 sec time delay. The specific limit at 26.5 of vacuum is 5 minutes; plausible as 90 seconds is the limit below 26 of vacuum.

C is wrong because closed is incorrect - see B.

D is correct because the anti-motoring relay actuates with a 5 sec time delay. The motoring limit with the exciter field breaker closed is 5 minutes with vacuum at 26.5.

Technical

References:

AOP-0002, MAIN TURBINE AND GENERATOR TRIPS, Rev 27 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(4)

Examination Outline Cross-Reference Level RO 262002 UPS (AC/DC) Tier # 2 Group # 1 2.4.31 Knowledge of annunciator alarms, K/A # 2.4.31 indications, or response procedures. Rating 4.2 Question 10 Which of the following conditions will cause the receipt of annunciator H13-P808/87A/A08, ENB-INV01B / 01B1 VITAL BUS INV TROUBLE?

1. Single Fan Failure

2. High DC Battery Voltage

3. Bypass Source Overvoltage

4. Transfer to Bypass Source

5. Inverter output off frequency A. 2 & 4 only B. 3 & 5 only C. 1, 3 & 4 only D. 1, 4 & 5 only Answer:

Explanation:

A is wrong because 2 and 3 do not cause alarms. Plausible as low battery voltage and bypass source undervoltage cause the alarm.

B is wrong see A & D C is wrong see A & D D is correct because a single fan failure, transfer to bypass source, and off frequency cause the alarm.

Technical

References:

R-STM-0300, REV 028 - AC Distribution ARP-808-87, P808-87 ALARM RESPONSE, Rev 24 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F4

Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 300000 Instrument Air System Tier # 2 Group # 1 Knowledge of electrical power supplies to the K/A # K2.01 following: Instrument air compressor Rating 2.8 Question 11 Which of the following provides power to instrument air compressor IAS-C2B?

A. NJS-SWG1E B. NJS-SWG1F C. NJS-SWG1G D. NJS-SWG1H Proposed Answer: D Explanation (Optional):

A. INCORRECT. This is the power supply to service air compressor SAS-C3A.

B. INCORRECT. This is the power supply to instrument air compressor IAS-C2C.

C. INCORRECT. This is the power supply to instrument air compressor IAS-C2A.

D. CORRECT. This is the power supply to instrument air compressor IAS-C2B.

Technical Reference(s): EE-001AC, Startup Electrical Distribution Chart, Rev 47 Proposed references to be provided to applicants during examination: None Learning Objective: RLP-STM-0121, Objective E Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge F3 10 CFR Part 55 Content: 55.41.7 Comments:

Examination Outline Cross-Reference Level RO 295037 SCRAM Condition Present and Tier # 1 Reactor Power Above APRM Downscale or Group # 1 Unknown K/A # EK1.07 Rating 3.4 Knowledge of the operational implications of the following concepts as they apply to SCRAM condition present and reactor power above APRM downscale or unknown:

Shutdown margin Question 12 The unit was at full power when an event occurred that required a SCRAM and it did not occur. As the ATC you placed the mode switch in shutdown, pushed the reactor SCRAM pushbuttons, and initiated ARI. Current plant conditions are:

• Reactor power is 10%

• several rods are not fully inserted What is the operational implication of this event because of the lack of adequate shutdown margin?

A. Only the Power and Level must be controlled differently than EOP-1 strategies with respect to shutdown margin in order to prevent damage to the core and the RPV B. Only the Power and Level must be controlled differently than EOP-1 strategies with respect to shutdown margin in order to prevent damage to the core only C. Power, Pressure, and Level must be controlled differently than EOP-1 strategies with respect to shutdown margin in order to prevent damage to the core and the RPV D. Power, Pressure, and Level must be controlled differently than EOP-1 strategies with respect to shutdown margin in order to prevent damage to the core only Answer: C Explanation:

A. Incorrect because all three legs must be controlled differently in EOP-1A versus EOP-1 to prevent damage to the RPV and core.

B. Incorrect because all three legs must be controlled differently in EOP-1A versus EOP-1 to prevent damage. Also the second part of this distracter does not include the RPV and is can be potentially damaged due to this event if not managed in accordance with EOP-1A.

C. Correct. All three legs must be controlled differently in EOP-1A versus EOP-1 to prevent damage to the RPV and core.

D. Incorrect First part is correct but the second part of this distracter does not include the RPV and is can be potentially damaged due to this event if not managed in accordance with EOP-1A.

Technical

References:

EPSTG-0002, page B-6-5 and B-6-6, B-5-14, revision 17.

References to be provided to applicants during exam: None.

Learning Objective: HLO-0513, Objective 4 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)8

Examination Outline Cross-Reference Level RO 295012 High Drywell Temperature Tier # 1 Group # 2 Ability to determine and/or interpret the K/A # AA2.02 following as they apply to high drywell Rating 3.9 temperature: AA2.02 Drywell pressure Question 13 Per the basis for EOP-1A, the Heat Capacity Temperature Limit (HCTL) is the highest suppression pool temperature that ensures RPV depressurization will NOT result in exceeding the ______ before _____.

A. Primary Containment Pressure Limit only; the rate of energy transfer from the RPV to the primary containment is within the capacity of the containment vent.

B. Primary Containment Pressure Limit only; suppression pool temperature impacts core heat removal.

C. Primary Containment Pressure Limit and maximum temperature capability of required containment equipment; suppression pool temperature impacts core heat removal.

D. Primary Containment Pressure Limit and maximum temperature capability of required containment equipment; the rate of energy transfer from the RPV to the primary containment is within the capacity of the containment vent.

Answer:

Explanation:

A is wrong because... the HCTL prevents exceeding containment pressure limit and temperature limits for required equipment. Plausible if applicant considers the primary goal of being within the capacity of the containment vent as being primarily pressure related.

B is wrong because the purpose of HCTL is to ensure the RPV energy release will be within the capacity of the containment vent. Plausible if applicant is concerned with heating up of suppression pool.

C is wrong because see B D is correct because the basis states that pressure and temperature limits are a concern and that the ultimate goal is remaining within the capacity of the containment vent.

Technical

References:

EOP-1A Basis for step RPA-4 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(8)

Examination Outline Cross-Reference Level RO 295001 Partial or Complete Loss of Forced Tier # 1 Core Flow Circulation Group # 1 K/A # AK3.02 Knowledge of the reasons for the following Rating 3.7 responses as they apply to partial or complete loss of forced core flow circulation: Reactor power response Question 14 According to AOP-0062, Jet Pump Failures, for a displaced jet pump mixer, core thermal power will ______ and flow in the other jet pump on the same riser will _____.

A. decrease; increase B. remain constant; decrease C. remain constant; increase D. decrease; decrease Answer:

Explanation:

A is wrong because flow in the other jet pump lowers. Plausible if applicant believes other jet pump flow will increase to compensate.

B is wrong because reactor power will decrease due to less core flow. Plausible if applicant believes that overall core flow remains unchanged.

C is wrong because see A & B.

D is correct because reactor power will decrease due to less core flow and the second jet pump flow will lower due to preferential flow through the failed jet pump.

Technical

References:

AOP-0062, Jet Pump Failures, Rev 002 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level RO 295029 High Suppression Pool Wtr Lvl Tier # 1 Group # 2 Knowledge of the interrelations between high K/A # EK2.08 suppression pool water level and the following: Rating 2.6 Drywell/suppression chamber ventilation Question 15 A LOCA is in progress due to an RPV leak. The crew enters EOP-2, Primary Containment Control. The following indications are observed in the control room:

• Containment pressure indicates 12 psig and increasing uncontrollably

• Suppression pool level indicates offscale high The operating crew should:

A. Secure normal containment venting to prevent damage to HVAC ductwork in the auxiliary building B. Terminate RPV injection from sources external to CTMT not needed to assure adequate core cooling or to shut down the reactor C. Emergency ventilate the containment to control pressure below 30 psig.

D. Determine containment pressure and suppression pool level using Enclosure 23.

Answer:

Explanation:

A is wrong because Enclosure 23 should be used to verify level and pressure. Plausible as this is an action to take for high containment pressure.

B is wrong because Enclosure 23 should be used to verify level and pressure. Plausible as this is an action to take for high SP level.

C is wrong because Enclosure 23 should be used to verify level and pressure. Plausible as this is an action to take for high containment pressure.

D is correct because with suppression pool level offscale high, Enclosure 23 should be used to calculate actual containment pressure and suppression pool level prior to taking any actions.

Technical

References:

EOP-2 Basis for SPL-4 and Caution #7 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level RO 295018 Partial or Total Loss of CCW Tier # 1 Group # 1 Knowledge of the reasons for the following K/A # AK3.04 responses as they apply to partial or complete Rating 3.3 loss of component cooling water: Starting standby pump Question 16 The Unit has experienced a loss of CCW pump and RPCCW pressure is lowering.

Current CCW pressure is 94 psig and lowering.

Based on the above plant conditions the standby CCW pump will start to prevent CCW pressure from lowering to _______ where ___________ will occur.

A. 65 psig, Division I and Division II of CCW Extreme Low Pressure isolations ONLY will occur B. 56 psig, Division I and Division II of CCW Extreme Low Pressure isolations ONLY will occur C. 65 psig, Division I of CCW Extreme Low Pressure Isolation ONLY will occur D. 56psig, Division II of CCW Extreme Low Pressure Isolation ONLY will occur Answer:

Explanation:

A. Incorrect, CCW extreme low pressure is actuated by 56 psig CCW pressure B. Correct C. Incorrect, CCW extreme low pressure is actuated by 56 psig CCW pressure and both division will actuate at this pressure D. Incorrect, Both division of CCW extreme low pressure will actuate that the given pressure Technical

References:

AOP-0011 References to be provided to applicants during exam: None.

Learning Objective: Objective 5 of RLP-OPS-AOP11.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H2

10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 295019 Partial or Total Loss of Inst. Air Tier # 1 Group # 1 Ability to operate and/or monitor the following K/A # AA1.01 as they apply to partial or complete loss of Rating 3.5 instrument air: Backup air supply Question 17 The plant is at 100% power when pressure in the instrument air system begins to lower.

What pressure would you expect to see SAS-AOV134, Instrument Air Header Cross-Tie Valve begin to open?

A. 120 psig B. 117 psig C. 114 psig D. 113 psig Answer: D Explanation: 113 psig is the pressure that open the cross connect valve. All other values are associated with starting of system air compressors.

Technical

References:

R-STM-0121, Plant Air Systems, Rev. 16, p. 26 References to be provided to applicants during exam: None.

Learning Objective: State how the compressed air subsystems can be cross-connected and state what happens if Instrument Air pressure begins to lower.

List the automatic functions and interlocks for each of the following plant air system components: SAS-IAS Cross Tie Valve (SAS-AOV134).

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 209001 LPCS Tier # 2 Group # 1 Knowledge of the operational implications of K/A # K5.04 the following concepts as they apply to LOW Rating 2.8 PRESSURE CORE SPRAY SYSTEM: K5.04 Heat removal (transfer) mechanisms Question 18 While in the EOP network the following conditions exist:

RPV level is -195 inches and stable RPV pressure is 425 psig LPCS flow rate is 5010 gpm What is the PRIMARY method of core cooling?

A. Core Submergence B. Spray Cooling C. Steam Cooling with Injection D. Steam Cooling without Injection Answer: B Explanation:

A is wrong because according to EOP-1 core submergence is the primary method above -

162 inches B is correct because according to EOP-1 spray cooling is assured with LPCS flow greater than 5000 gpm and RPV level above -211 inches C is wrong because steam cooling with injection is assured greater than -187 inches D is wrong because steam cooling without injection is assured greater than -200 inches.

Technical

References:

EPSTG-02, EOP Bases References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis

10CFR Part 55 Content: 55.41(b)(14)

Examination Outline Cross-Reference Level RO 295031 Reactor Low Water Level Tier # 1 Group # 1 Ability to determine operability and/or K/A # 2.2.37 availability of safety related equipment. Rating 3.6 Question 19

• An electrical fault occurs which requires removing power to power supply EHS-MCC2J for repairs.

• A LOCA occurs during repair activities, requiring implementation of EOP-1, RPV Control Which of the Preferred Injection Systems, listed in Table L-1, would not be available?

A. RHR B Train in LPCI Mode B. LPCS C. HPCS D. RHR C Train in LPCI Mode Answer: B Explanation:

A is wrong because the source bus for RHR pump B and its related MCC loads is ENS-SWG1B.

B is correct because power supply EHS-MCC2J (source bus ENS-SWG1A) powers E21-MOVF005, LPCS pump discharge valve. This valve is normally closed, so electrical power is needed to open it to provide injection when if reactor vessel level reaches Level 1 (-143).

C is wrong because the source bus for HPCS and its MCC loads is source bus SWGR E22-S004.

D is wrong because the source bus for RHR pump C and its related MCC loads is ENS-SWG1B.

Technical

References:

R-STM-0205, Low Pressure Core Spray System, Revision 6 R-STM-0204, Residual Heat Removal System, Revision 12 R-STM-0203, High Pressure Core Spray System (HPCS), Revision 8 R-STM-0300, AC Distribution, Revision 28 EOP-1, RPV Control, River Bend Station, Revision 27 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis X 10CFR Part 55 Content: 55.41(b)(7) 55.43

Examination Outline Cross-Reference Level RO 215003 IRM Tier # 2 Group # 1 Ability to monitor automatic operations of the K/A # A3.02 INTERMEDIATE RANGE MONITOR (IRM) Rating 3.3 SYSTEM including: A3.02 Annunciator and alarm signals Question 20

• The plant is conducting a reactor startup, with the Reactor Mode Selector Switch is in the Startup position.

• Power is available to all neutron monitoring instrumentation and drive control.

• Alarm Window C01, CONTROL ROD WITHDRAWAL BLOCK, Alarm Panel P680-07, is energized Which of these indications would be consistent with a control rod block condition ONLY while power is being monitored in the intermediate range?

A. IRM F IN light deenergized on the drive control panel B. Digital Recorder C51-R603A, red pen reads 122 on the 0-125 scale of Range 4 C. INOP light on IRM D Drawer is energized D. Digital Recorder C51-R603B, red pen reads 1 on the 0-40 scale of Range 1 Answer: A Explanation:

A is correct because the combined indication is that of an IRM detector in the wrong position.

R-STM-0503, Page 35 of 112, says that the indications of this condition are the IN light on the IRM Drive Select Matrix and a rod block annunciator (with drive power available). This only causes a rod withdrawal block.

B is wrong because this condition would cause both a rod withdrawal block and a reactor scram signal. The rod block and scram set points are 108/125 and 120/125, respectively.

See Pages 34 and 36 of R-STM-0503.

C is wrong because an Inoperative IRM will cause both a rod withdrawal block and a reactor scram signal. See Pages 35-36 of R-STM-0503.

D is wrong because IRMs on Range 1 bypasses the IRM Downscale rod withdrawal block (

5/125 scale). See Page 34 of R-STM-503. Therefore, this condition would not support the existence of the rod block indication provided in the questions initial conditions.

NOTE: The ARP does NOT list the IRM not being in the full IN position as being a reason for the rod block annunicator. This conflicts with the STM document. This needs to be verified, and the question revised as needed.

Technical

References:

System Training Manual R-STM-0503, Neutron Monitoring Instruments System, Revision 9 Procedure ARP-680-07, P608-07 Alarm Response, Revision 35 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(7) s

Examination Outline Cross-Reference Level RO 259002 Reactor Water Level Control Tier # 2 Group # 1 Ability to predict and/or monitor changes in K/A # A1.03 parameters associated with operating the Rating 3.8 REACTOR WATER LEVEL CONTROL SYSTEM controls including: Reactor power Question 21 The reactor is operating at 85% power. RPV water level is being controlled with the Feedwater Regulating Valve Master Flow Controller in MANUAL due to recent maintenance which has been completed.

Reactor water level is lowering because of an inexperienced operator on the feedwater equipment.

When happens when RPV water level drops to Level 4?

What happens when RPV level drops to Level 3 (after the SCRAM and Recirc Pumps downshift to slow speed)?

A. A Recirc FCV Runback will occur.

Master Flow Controller Setpoint Setdown is initiated.

B. A Recirc FCV Runback will NOT occur.

Master Flow Controller Setpoint Setdown is not initiated.

C. A Recirc FCV Runback will occur.

Master Flow Controller Setpoint Setdown is not initiated.

D. A Recirc FCV Runback will NOT occur.

Master Flow Controller Setpoint Setdown is initiated.

Answer: B Explanation:

A is wrong because 3 feed pumps are running, so the Recirc FCV Runback will not occur.

Also since the Master Flow Controller is in manual, Setpoint Setdown is not initiated.

B is correct.

C is wrong because 3 feed pumps are running, so the Recirc FCV Runback will not occur.

D is wrong because since the Master Flow Controller is in manual, Setpoint Setdown is not initiated.

Technical

References:

R-STM-0107, Rev 27, REACTOR FEEDWATER AND LEVEL CONTROL SYSTEMS References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0107 Obj # B3, H5, L4

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 215005 APRM / LPRM Tier # 2 Group # 1 2.4.34 Knowledge of RO tasks performed K/A # 2.4.34 outside the main control room during an Rating 4.2 emergency and the resultant operational effects.

Question 22

• The Main Control Room is evacuated due to a fire.

• The reactor is manually scrammed from 100% power prior to evacuation. Control rod positions are unknown.

• The ATC proceeds to the Division 1 Remote Shutdown Panel to implement AOP-0031 Attachment 12, Shutdown Outside Main Control Room: ATC Operator Actions.

After 6 minutes, the ATC observes 1 SRV full open on the RSP.

1) What APRM-equivalent value of thermal power does this indicate?

2) What is the source of this power?

A. 1) 3%

2) decay heat only B. 1) 7%

2) decay heat only C. 1) 3%

2) decay heat plus fission heat D. 1) 7%

2) decay heat plus fission heat Answer: D Explanation:

There is no APRM/LPRM indication of power at the Div I Remote Shutdown Panel. Reactor power must be estimated by the number of SRVs open.

Each SRV full open is 7% power. Decay heat 6 minutes after a SCRAM is approximately 3%

thermal power which is indicated by one SRV cycling.

IF Reactor power is greater than 3%, 6 minutes after the SCRAM, THEN enter EOP-0001, Emergency Operating Procedure - RPV Control.

To answer this question, the applicant must understand how to approximate reactor power from SRV position at the RSP, know the expected value of thermal power post-scram, and recognize that a positive deviation from this expected value of thermal power must be an indication of continued fission above the point of adding heat. This is knowledge required for an RO to effectively perform his functions at the RSP.

A is wrong because 3% power would be indicated by one cycling SRV, and the RSP indicates that there is continuing fission above POAH in addition to decay heat.

B is wrong because one SRV full open 6 minutes post-SCRAM indicates that there is continuing fission above POAH in addition to decay heat.

C is wrong because 3% power would be indicated by one cycling SRV.

D is correct because one SRV full open is equivalent to 7% power, and since decay heat is expected to be 3% thermal power post-SCRAM, the increase is due to fission above POAH.

Technical

References:

AOP-0031, Shutdown Outside Main Control Room, Attachment 12, Revision 323.

References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(6)

Developer Working Notes:

Or 6 minutes following the SCRAM, what Remote Shutdown Panel indication of power should the RO expect to observe, equivalent to 3% APRM indication?

Or Several rods fail to insert, resulting in power stabilizing at 7% APRM indication. What indication of reactor power should the RO expect to observe at the RSP?

What action should be taken after 6 minutes?

After 6 minutes, the RO observes 1 SRV full open on the RSP. 1) What APRM-equivalent value of reactor power is this equivalent to? 2) What action should be taken?

Examination Outline Cross-Reference Level RO 204000 Reactor Water Cleanup System Tier # 2 Group # 2 Knowledge of Reactor Water Cleanup System K/A # K4.03 design feature(s) and/or interlocks which Rating 2.9 provide for the following: Over temperature protection for system components.

Question 23 Temperature element G33-N007, Non-Regenerative Heat Exchanger Outlet temperature, is reading 142°F. Which of the following automatic actions has occurred due to this temperature?

A. G33-F044, RWCU F/D Bypass Valve, automatically opened B. G33-F104, RWCU HX/DEMIN Bypass Valve, automatically opened C. G33-F004, RWCU Pump Outboard Suction Valve, automatically closed D. G33-F054, RWCU Pump Outboard Discharge Valve, automatically closed Proposed Answer: C Explanation (Optional):

A. INCORRECT. G33-F044, RWCU F/D Bypass Valve, is used to throttle flow around the filter demineralizers, and has no automatic function.

B. INCORRECT. G33-F104, RWCU HX/Demin Bypass Valve, is provided to allow for throttling flow around both the RHX and NRHX.

C. CORRECT. NRHX Outlet Temperature, or F/D Inlet Temperature, is interlocked with G33-F004, RWCU Pump Outboard Suction Valve, such that if temperature reaches 140°F, G33-F004 isolates.

D. INCORRECT. G33-F054, RWCU Pump Outboard Discharge Valve, RWCU Pump Room ambient tempratures.

Technical Reference(s): STM-0601, Reactor Water Cleanup (RWCU) System, Revision 8 (Attach if not previously provided, including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: RLP-STM-0601, Objective D Question Source: Modified Question History: Last NRC Exam Grand Gulf 2010-06 Question Cognitive Level: Memory or Fundamental Knowledge F3 10 CFR Part 55 Content: 55.41.7 Comments:

Original Bank Question, Grand Gulf, 2010-06 Exam

Examination Outline Cross-Reference Level RO 295002 Loss of Main Condenser Vac Tier # 1 Group # 2 Knowledge of the reasons for the following K/A # AK3.05 responses as they apply to loss of main Rating 3.4 condenser vacuum: Main steam isolation valve: Plant-Specific Question 24 The reactor plant is operating at 100% power. Changes in the plant result in the following indications:

• Main steam line pressure is 1085 psig

• Reactor vessel level is 30

• Main condenser vacuum is 8 Hg

• Main steam line tunnel temperature is 1600F Based on these conditions, the Main Steam Isolation Valves (MSIVs) automatically close. In this case, what is the reason for this protective action?

A. To isolate a leak in the main steam lines B. To prevent a rapid depressurization of the reactor pressure vessel C. To prevent exceeding reactor pressure vessel cooldown rate limitations D. To isolate or prevent a leak in the main condenser Answer: D Explanation:

A is wrong because this is the reason for an automatic closure of MSIVs for main steam line tunnel temperature. The set point for this is 1730F.

B is wrong because this is one of the reasons for an automatic MSIV closure for main steam line low pressure. The set point for this is 849 psig.

C is wrong because this is the other reason for the automatic MSIV closure for main steam line low pressure.

D is correct because main condenser vacuum 8.5 Hg automatically closes the MSIVs. The reason for this automatic capability is to protect against leaks in the main condenser. See R-STM-0109, Page 37 of 95. If there isnt a leak, the action is taken to prevent one as well -

see R-STM-0058, Page 10 of 63.

Technical

References:

System Training Manual R-STM-0109, Main Steam System, Revision 15 System Training Manual R-STM-0051, Nuclear Boiler Instrumentation, Revision 6 System Training Manual R-STM-0058, Containment and Reactor Vessel Isolation Control System (CRVICS), Revision 10 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level RO Tier # 3 Group #

Knowledge of the process for controlling K/A # 2.2.14 equipment configuration or status. Rating 3.9 Question 25 A pump in the plant has developed a greater than normal seal leak allowed by operator rounds.

The operator secured the pump and the started the standby pump as directed.

Which of the following describes the appropriate control method for the status of the secured pump?

A. Place a Test and Maintenance Tag on the switch to allow maintenance activities.

B. Place a Lockout Device on the switch to prevent use of this equipment.

C. Place a Danger Tag on the switch to prevent use of this equipment.

D. Place a Caution Tag on the switch that documents the condition of the equipment.

Proposed Answer: D Explanation (Optional):

A. INCORRECT. This answer is plausible because a T&M tag would be used for a component that would need to be manipulated during maintenance activities, such as determining the leak location, but is incorrect because a T&M tag permits operation of the equipment by only authorized persons signed on to the tagout.

B. INCORRECT. This answer is plausible because during maintenance activities a lockout device would be used however the device would be installed on the pump breaker.

C. INCORRECT. This answer is plausible because the Danger hold tag would be used to place the equipment in a safe condition if maintenance were to be performed. When maintenance is to be performed, the Caution tag would be replaced by a Danger tag.

D. CORRECT. Caution tags provide precautions or special instructions that relate to unusual or out-of-normal conditions.

Technical Reference(s): EN-OP-102, Revision 18, page 5 and 60 (Attach if not previously provided, including version/revision number)

Proposed references to be provided to applicants during examination: None Learning Objective: RLP-HLO-0216 Objective G Question Source: Bank Question History: Last NRC Exam 2015 Question Cognitive Level: Memory or Fundamental Knowledge 2 10 CFR Part 55 Content: 55.41.10

Comments:

Examination Outline Cross-Reference Level RO 295006 SCRAM Tier # 1 Group # 1 Knowledge of the operational implications of K/A # AK1.01 the following concepts as they apply to Rating 3.7 SCRAM: AK1.01 Decay heat generation and removal Question 26 The plant has been operating at full power for 9 months when an electrical transient causes all of the turbine control valves to fully open. The reactor automatically Scrams and no operator action is taken.

One minute following the Scram, decay heat is being removed by ___(1)____ and __(2)___

are also available to assist in removing decay heat.

A. 1) SRVs in Low-Low Set Mode

2) RWCU System Components B. 1) SRVs in Relief Mode

2) RWCU System Components C. 1) SRVs in Low-Low Set Mode

2) Main Turbine Bypass Valves D. 1) SRVs in Relief Mode

2) Main Turbine Bypass Valves Answer: A Explanation:

A is correct. On a full open of all control valves the reactor scrams on low steam pressure and the MSIVs go closed. The bypass valves would not be available to control decay heat but RWCU is available to help with removing decay heat after MSIV closure.The SRVs (in auto) would initially control in relief mode but would transition to low low set mode, which makes A correct.

B is incorrect because SRVs would control in low-low set mode.

C is wrong because bypass valves are not available once MSIVs go closed.

D is correct because SRVs control in low-low set mode and bypass valves are not available Technical

References:

R-STM-0109, revision 15.

References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0109, Obj E Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO 216000 Nuclear Boiler Instrumentation Tier # 2 Group # 2 Knowledge of the physical connections and/or K/A # K1.06 cause/effect relationships between Nuclear Rating 3.9 Boiler Instrumentation and the following: Low pressure core spray Question 27 Which of the following trip units are associated with the initiation of Low Pressure Core Spray?

A. Level 1 trip B. Level 2 trip C. Level 3 trip D. Level 4 trip Proposed Answer: A Explanation (Optional):

A. CORRECT. Low Pressure Core Spray is associated with a Level 1 trip.

B. INCORRECT. RCIC y is associated with a Level 2 trip.

C. INCORRECT. SCRAM is associated with a Level 3 trip.

D. INCORRECT. Feed pump is associated with a Level 4 trip.

Technical Reference(s):

STM-0051, Nuclear Boiler Instrumentation, Revision 6 Proposed references to be provided to applicants during examination: None Learning Objective: RLP-STM-0051, Objective M Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Fundamental F3 10 CFR Part 55 Content: 55.41.2 to 41.9 Comments:

Examination Outline Cross-Reference Level RO 295027 High Containment Temperature Tier # 1 Group # 1 Knowledge of the reasons for the following K/A # EK3.01 responses as they apply to high containment Rating 3.7 temperature (Mark III containment only):

Emergency depressurization: Mark-III Question 28 In accordance with EOP-2 and its associated bases for step CT-6, a containment temperature of ___(1)_____ requires Emergency Depressurization ___(2)_____.

A. 1) 185°F

2) In order to extend equipment operability within containment for as long as possible during an event B. 1) 145°F

2) In order to extend equipment operability within containment for as long as possible during an event C. 1) 185°F

2) While the rate of energy transfer from the RPV to containment is less than the capacity of the containment coolers/vents during an event D. 1) 145°F

2) While the rate of energy transfer from the RPV to containment is less than the capacity of the containment coolers/vents during an event Answer: A Explanation:

A is correct in accordance with EOP-2 and its bases, step CT-6, an ED is required at 185F to terminate, or reduce as much as possible, any continued containment temperature increase and thereby maintain equipment operability for as long as possible.

B is incorrect because the 145F temperature is the DW cooling SP closest to this SP but is not 185F and is not correct. The second part is correct.

C is incorrect because although the temperature is correct the second part is wrong and these words are part of the basis for step SPT-6 to ED when HCTL is exceeded.

D is incorrect because both parts are incorrect.

Technical

References:

EOP-2 and EOP-2 bases, step CT-6, page B-8-9, Revision 17 References to be provided to applicants during exam: None.

Learning Objective: HLO-514 Obj. 5 Question Source: Bank # X (2008R audit exam)

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)9

Examination Outline Cross-Reference Level RO 211000 Standby Liquid Control System Tier # 2 Group # 1 Ability to predict and/or monitor changes in K/A # A1.09 parameters associated with operating the Rating 4.0 STANDBY LIQUID CONTROL SYSTEM controls including: SBLC system lineup Question 29 The plant is operating at 100% power, and the operators are preparing to run SLC Pump A with a suction from the test tank. Test tank outlet valve, C41-F031, has been opened, but the pump has yet to be started.

Subsequently, a reactor SCRAM occurs with an ATWS. The CRS has directed injection with Standby Liquid Control using SLC Pump B. The operator takes the pump switch for SLC Pump B to RUN.

What will be the position of the Tank Outlet Valve, C41-F001B, and the Squib Continuity Valve, C41-F004B?

A. C41-F001B OPEN, C41-F004B OPEN B. C41-F001B OPEN, C41-F004B CLOSED C. C41-F001B CLOSED, C41-F004B OPEN D. C41-F001B CLOSED, C41-F004B CLOSED Proposed Answer: C Explanation (Optional):

A. INCORRECT. C41-F001B is interlocked to C41-F031, and will not open unless C41-F031 is closed. C41-F004B will open when the pump switch is taken to run.

B. INCORRECT. C41-F001B is interlocked to C41-F031, and will not open unless C41-F031 is closed. C41-F004B will open when the pump switch is taken to run.

C. CORRECT. C41-F001B is interlocked to C41-F031, and will stay closed since C41-F031 is open. C41-F004B will open when the pump switch is taken to run.

D. INCORRECT. C41-F001B is interlocked to C41-F031, and will stay closed since C41-F031 is open. C41-F004B will open when the pump switch is taken to run.

Technical Reference(s):

STM-0201, Standby Liquid Control, Revision 4 Proposed references to be provided to applicants during examination: None Learning Objective: RLP-STM-0201, Objective F (a)

Question Source: New Question History: N/A Question Cognitive Level: Comprehension or Analysis H3

10 CFR Part 55 Content: 55.41.5 Comments:

Examination Outline Cross-Reference Level RO 223002 PCIS/Nuclear Steam Supply Shutoff Tier # 2 Group # 1 Knowledge of the effect that a loss or K/A # K3.06 malfunction of the PRIMARY CONTAINMENT Rating 2.8 ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF will have on following:

Turbine building radiation Question 30 A failure of the Containment and Reactor Vessel Isolation Control System (CRVICS) to initiate an isolation signal for a steam line break in the Turbine Building would have the following effect?

A. Cause Standby Gas Treatment System to initiate B. Raise area radiation levels C. Cause cavitation and damage to an operating RHR pump D. Fail to isolate Turbine Building Chilled Water to the Containment Unit Coolers Answer: B Explanation:

A is wrong because a) the Standby Gas Treatment Systems actuation is affected by the Balance of Plant (BOP) LOCA Isolation Logic in the CRVICS, and b) if the portion of CRVICS that affects SGTS were to fail to actuate, it would result in SGTS NOT starting. See R-STM-0058, Pages 18 and 41 of 63.

B is correct because a failure to isolate the MSIVs and Main Steam Line drains, which protect against a steam line break, would result in Turbine Building radiation levels increasing with a break in that area. See R-STM-0058, Page 41 of 63.

C is wrong because a) RHR is affected by the RHR Isolation Logic portion of CRVICS, and b) this is the result if there is a leak in the shutdown cooling line that the associated isolation logic fails to address. See R-STM-0058, Pages 29 and 42 of 63.

D is wrong because the isolation of Turbine Building Chilled Water is affected by the BOP LOCA Initiation Logic in the CRVICS. See R-STM-0058, Pages 17 and 42 of 63.

Technical

References:

System Training Manual R-STM-0058, Containment and Reactor Vessel Isolation Control System (CRVICS), Revision 10 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content: 55.41(b) (7)

Examination Outline Cross-Reference Level RO 212000 RPS Tier # 2 Group # 1 Ability to (a) predict the impacts of the K/A # A2.16 following on the REACTOR PROTECTION Rating 4.0 SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Changing mode switch position Question 31 A reactor startup is in progress, using procedure GOP-0001 Reactor power is 13% and stable Reactor System Mode Switch is in the RUN position Intermediate Range Monitors (IRMs) are still inserted, with all on Range 9 with 110/125 of scale a) If an operator took the Reactor System Mode Switch to the START & HOT STBY position, what effect would it have on the Reactor Protection System?

b) Based on these conditions, what procedural entry is necessary?

A. a) Initiates a reactor scram a) AOP-0001, Reactor Scram B. a) Initiates a reactor scram b) GOP-0002, Power Decrease/Plant Shutdown C. a) Bypasses Reactor Water Level - High signal b) AOP-0001, Reactor Scram D. a) Bypasses Reactor Water Level - High signal b) GOP-0002, Power Decrease/Plant Shutdown Answer: D Explanation:

A is wrong. This is plausible if the applicant believes that any of the reactor scram set points in effect have been exceeded.

B is wrong. There is no reactor scram. Even if one was initiated, after AOP-0001 is entered and completed, it directs entry afterwards into GOP-0003, Scram Recovery.

C is wrong. The Reactor Water Level - High signal is bypassed. However, nothing indicated in plant conditions would cause a reactor scram, and there isnt a condition that would cause an operator to initiate a reactor scram.

D is correct. Per R-STM-0508, when the Reactor System Mode Switch is in the START &

HOT STBY position, the MSIV Closure and Reactor Water Level - High scram signals are bypassed. Also, the APRM Neutron Flux - High scram set point is reduced to 15% power.

The IRMs are still inserted, so their scram set point is 120/125 of scale on its ranges (R-STM-0503). Based on these conditions, there is no signal that would cause a reactor scram, and there isnt a condition that should cause an operator to initiate a reactor scram. The purpose of GOP-0002 is To provide guidelines to reduce power from any point to any standby or shutdown mode. Therefore, this procedure would be entered to address the

situation.

Technical

References:

R-STM-0500, Rod Control and Information System, Revision 4 R-STM-0503, Neutron Monitor Instruments, Revision 9 R-STM-0508, Reactor Protection System, Revision 6 GOP-0001, Plant Startup, Revision 86 GOP-0003, Scram Recovery, Revision 27 GOP-0002, Power Decrease/Plant Shutdown, Revision 73 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(5)

Notes:

1) Searched administrative procedures for direction that states that AOP/EOP entry is prioritized above other procedures, but couldnt find it. This may be needed.

2) Based on review of GOP-0001, the plant is around Page 50 of 103 in the procedure on the startup defined in the initial conditions.

3) Need to verify with plant staff if the IRMs would be on Range 9 or 10 and what scale they would be on in this scenario. Ensure that the indications are realistic.

Examination Outline Cross-Reference Level RO 209002 HPCS Tier # 2 Group # 1 Knowledge of the effect that a loss or K/A # K6.02 malfunction of the following will have on the Rating 3.4 HIGH PRESSURE CORE SPRAY SYSTEM (HPCS): Condensate Storage Tank Water Level Question 32 A stroke test of MOVF001 HPCS CST Suction Valve was in progress when the valve became mechanically bound on its closed seat. While the valve is bound in the closed position a large loss of coolant accident occurs that requires HPCS to actuate.

With NO operator action the HPCS system will start and ____________ provide cooling water flow to the RPV __________.

A. Will, when MOVF015, HPCS Suppression Pool Suction Valve automatically opens at a < 2.4 feet CST level B. Will not, because with MOVF001 HPCS CST Suction valve closed CST level will not reach the actuation set point for MOVF015 HPCS Suppression Pool Suction Valve C. Will, MOVF015, HPCS Suppression Pool Suction Valve automatically opens due to high suppression pool level D. Will not, because without either MOVF001 or MOVF015 open the HPCS system automatically trips off line Answer: B Explanation:

A. Incorrect while 2.4 feet in the CST is a signal to open MOV015 the level will not be reached due to no flow through the normal suction valve from the CST B. Correct C. Incorrect, High suppression pool level is a automatic open set point for MOVF015 it would not be reached will a large break loss of coolant accident occurring D. Incorrect, There is not automatic trip of the HPCS pump based on these valve positions Technical

References:

R-STM-0203 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(7)

From Lesson plan

6. HPCS CST Suction Valve (E22-MOVF001) a) Discussion The HPCS CST Suction Valve (E22-MOVF001) is a normally opened 16" gate valve that automatically opens to provide suction from the CST upon receipt of a HPCS initiation signal. In AUTO the valve opens upon HPCS initiation provided that the Suppression Pool Suction Valve MOVF015 is not fully open. MOVF001 is slaved to MOVF015 such that if MOVF001 is open and MOVF015 subsequently opens, MOVF001 starts to close when MOVF015 reaches its full open position.

Auto-opens if there is a LOCA signal and E22-F015 is not full open. Cannot be manually closed until the LOCA signal is reset. Cannot be manually opened unless E22-F015 is not full open. Auto-closes whenever E22-F015 is full open.

HPCS Suppression Pool Suction Valve (E22-MOVF015) a) Discussion HPCS Suppression Pool Suction Valve (E22-MOVF015) is a normally closed 20" gate valve. Upon the receipt of either a low CST level (98'6" mean sea level or MSL,

< 2.4'CST level) or a high suppression pool level (20'4") signal, MOVF015 opens to provide the alternate suction supply to the HPCS pump.

Examination Outline Cross-Reference Level RO 239002 SRVs Tier # 2 Group # 1 Knowledge of RELIEF/SAFETY VALVES K/A # K4.01 design feature(s) and/or interlocks which Rating 3.9 provide for the following: Insures that only one or two SRVs reopen following the initial portion of a reactor isolation event (LLS logic).

Question 33 The design feature of Low-Low Set mode for the Safety Relief Valves is that the designated SRVs will _____(1)_____ in order to ____(2)_____.

A. (1) open earlier and stay open longer (2) reduce cyclical stresses on the RPV B. (1) open earlier and stay open longer (2) reduce cyclical stresses on the containment C. (1) open later and stay open shorter (2) reduce cyclical stresses on the RPV D. (1) open later and stay open shorter (2) reduce cyclical stresses on the containment Answer: B Explanation:

A is wrong because the second part is wrong, it is to reduce cyclical stresses on containment not the RPV.

B is correct, to open earlier and stay open longer to reduce stresses on containment C is wrong because both aspects are wrong.

D is wrong because the first part is wrong.

Technical

References:

R-STM-0109, Rev. 15, page 10.

References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis

10CFR Part 55 Content: 55.41(b)7 Examination Outline Cross-Reference Level RO 263000 DC Electrical Distribution Tier # 2 Group # 1 Knowledge of the physical connections and/or K/A # K1.01 cause/effect relationships between D.C. Rating 3.3 ELECTRICAL DISTRIBUTION and the following: A.C. electrical distribution Question 34 BYS-CHGR1D (BACKUP BATTERY CHARGER) is being placed IN service. What is the normal power supply for this battery charger AND which breaker should be closed first (AC or DC)?

A. EJS-SWG2B AC Breaker B. NJS-SWG1U1 DC Breaker C. EJS-SWG2B DC Breaker D. NJS-SWG1U1 AC Breaker Answer: B Explanation:

A is wrong because EJS-SWG2B is the power supply for battery charger IHS-CHGR1D. AC breaker is also incorrect. When placing a battery charger in service the DC breaker must be closed first, then the AC.

B is correct.

C is wrong because EJS-SWG2B is the wrong power supply. The DC breaker is correct.

D is wrong because closing the AC breaker first is incorrect.

Technical

References:

R-STM-0305, Rev. 7, DC Distribution References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)7 F. PRECAUTIONS AND LIMITATIONS

NOTE: Refer to SOP-0049, 125 VDC System for a complete list of all applicable Precautions and Limitations.

1. Battery room HVAC System should always be in operation when batteries are in service.

When entering battery rooms, caution should be used. Never smoke in or around battery rooms. Never use open flame in or around battery rooms. Use protective clothing when working on or around batteries.

2. NPS-SWG1C/D control power can be supplied from either BYS-PNL02A1 or BYSPNL02B1 via BYS-TRS1A or BYS-TRS1B. These transfer switches are a make-before -

break type switch, therefore a loss of control power should not occur . However when transferring control power supplies between BYS-TRS1A and BYS-TRS1B load transients on BYS-SWG01A or BYS-SWG01B such as cycling breakers supplied control power by these buses should be avoided.

3. When placing battery chargers in service, the DC breaker must be closed first, then close the AC breaker. When removing battery chargers from service, the AC breaker must be opened first, then open the DC breaker. If this procedure is not followed, it is possible to damage the rectifier stack or blow the anode fuses.

TABLE 2 POWER SUPPLIES EQUIPMENT DESCRIPTION POWER SUPPLY ENB-CHGR1A BATTERY CHARGER EJS-SWG1A ENB-CHGR1B BATTERY CHARGER EJS-SWG1B E22-S001CGR BATTERY CHARGER E22-S002 BYS-CHGR1A BATTERY CHARGER EJS-SWG1A BYS-CHGR1B BATTERY CHARGER EJS-SWG1B IHS-CHGR1D BATTERY CHARGER EJS-SWG2B BYS-CHGR06 BATTERY CHARGER NHS-MCC9K BYS-CHGR1C BATTERY CHARGER NHS-MCC11A BYS-CHGR04 BATTERY CHARGER NHS-MCC103A BXY-CHGR1 BATTERY CHARGER SCA-PNL12A1 BYS-CHGR1D BACKUP BATTERY NJS-SWG1U1 OR CHARGER BYS-EG1(SBO D/G)

Examination Outline Cross-Reference Level RO 295017 High Off-site Release Rate Tier # 1 Group # 2 Knowledge of the operational implications of K/A # AK1.02 the following concepts as they apply to high Rating 3.8 off-site release rate: Protection of the general public.

Question 35 Due to a steam leak, the Main Steam Line Tunnel area temperatures are all between 160°F and 170°F. All automatic isolations have occurred as designed. Because of the leak location and isolation actions, NO LOCA signal occurred from high drywell pressure or RPV low level.

An ALERT has been declared based on offsite release rates.

Which one of the following will reduce dose to the public by reducing the UNMONITORED release rate?

A. Shutdown the Turbine Building Ventilation System if operating.

B. Shutdown the Radwaste Building Ventilation System if operating.

C. Start the Turbine Building Ventilation System if NOT operating.

D. Start the Fuel Building Charcoal Filtration trains if NOT operating.

Answer: C Explanation:

A is wrong per RR-1 of EOP-3 for high dose rate associated with ALERT or higher, the turbine building vent system must be restarted if not running, so C is correct and the others are wrong.

B is wrong (see A above)

C is correct because this is the directed ventilation system and it is directed to restart it if not running.

D is wrong (see A above).

Technical

References:

EOP-3, RR-2, revision 17 EPSTG-2, Page B-10-4, revision 17 References to be provided to applicants during exam: None.

Learning Objective: HLO-515 OBJ- 6 Question Source: Bank # X (note changes; attach parent) Modified Bank #

New

Question History: Last NRC Exam 2003 Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)13

1 PROCEDUREÇ (EIP-2-001)

NOTE The assessment, classification, and declaration of an emergency condition is expected to be completed within 15 minutes after the availability of indications (i.e. plant instrumentation, plant alarms, computer displays, or incoming verbal reports) to plant operators that an EAL has been exceeded 1.1 Anytime an event occurs that has the potential of causing or resulting in a hazard to personnel, onsite or offsite, the Emergency Director:

1.1.1. Should review INITIATING CONDITIONs and EALs to determine if the event should be classified as an emergency.

1.1.2. Shall classify the emergency in accordance with this procedure and implement EIP-2-002, Classification Actions, if criteria are met.

1.2 River Bend Station Senior Management or designated alternate shall:

1.2.1. Provide assistance to the OSM, as requested, if the emergency is classified as an Unusual Event (NOUE).

1.2.2. Relieve the OSM of the responsibilities of Emergency Director as soon as practical for an ALERT or higher classification and implement applicable EIP procedures.

1.2.3. The Emergency Director will review this procedure and upgrade the emergency to a SITE AREA EMERGENCY or GENERAL EMERGENCY when warranted.

2 GENERAL (EIP-2-007) 2.1 The following table is a guideline for specific Protective Action Recommendations (PARs) (see Attachment 1).

Whole Body Thyroid Total Effective Committed Protective Action Dose Equivalent Dose Equivalent to be Recommended

< 1 rem OR < 5 rem No specific actions for the general public

> 1 rem OR > 5 rem Evacuate area unless constraints make evacuation impractical

> 25 rem Consider administration of stable iodine for emergency workers 2.2 The authority and responsibility for the selection and implementation of offsite response options rests fully with the appropriate State and local authorities. River Bend Station has no authority with respect to imposing protective response options beyond the boundaries of the River Bend Site.

2.3 Protective Action Recommendations are based on projected radiation exposure. State and local authorities may take into consideration ambient meteorology and duration of release.

Evacuation times and degree of protection afforded by local residential units are considered by the State as appropriate when considering sheltering in lieu of evacuation.

Examination Outline Cross-Reference Level RO 295016 Control Room Abandonment Tier # 1 Group # 1 Knowledge of the interrelations between K/A # AK2.02 control room abandonment and the following: Rating 4.0 Local control stations: Plant-Specific Question 36 A fire in the main control room is in progress and the CRS has ordered it to be abandoned.

What is required to be performed of the Reactor Building Operator in less than 10 minutes of scramming the reactor?

A. Locally initiate a full NSSS Isolation at the RPS MG sets and EPA Breakers at CB 116.

B. Locally isolate all 10 Condensate Demineralizers.

C. Align for remote operation plant electrical and HVAC systems for plant cooldown D. Transfer power at E51-MOVF063 from Div. II to Div. I Alternate Power.

Answer: D Explanation:

A is wrong because this is what is required of the Unit Operator (UO) and to be accomplished within five minutes.

B is wrong because this is what is required of the Aux Control Room Operator and to be accomplished within three minutes.

C is wrong because this is what is required of the RB Operator after the first ten minutes..

D is correct.

Technical

References:

AOP-0031, Rev. 323, Shutdown From Outside the Main Control Room References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)10

PURPOSE/DISCUSSIONÇ (AOP-0031) 1.1 To provide instructions for the safe shutdown and cooldown of the Reactor following an event requiring evacuation of the Main Control Room (MCR).

1.2 This procedure is written to cover a full spectrum of possible events including a Control Room fire.

1.2.1. It is based on the assumption that Control Room operations will not be regained for the duration of the event or until cold shutdown conditions are established.

1.2.2. Implementation of this procedure provides for the successful mitigation of the consequences of a catastrophic Control Room fire coincident with or without a loss of offsite power.

1.2.3. This procedure has been written to ensure that all Control Room fire-induced LOCA pathways are closed within a suitable time frame so as to ensure 10CFR50 Appendix R requirements are not exceeded.

Additionally, this procedure, in conjunction with the design of the plant, is written to account for and mitigate the consequences of worst case spurious signals originating from a MCR catastrophic fire.

1.2.4. During a Control Room fire event, only Division 1 and 3 equipment is assumed to be operable, with the main focus on using Division 1 equipment to put the station in a safe shutdown condition.

1.3 This procedure interfaces with the General Operating Procedures (GOP),

Emergency Operating Procedures (EOP), Abnormal Operating Procedures (AOP),

and System Operating Procedures (SOP). These procedures are not written to be performed from the RSS panels, but are provided as a source of reference if needed.

1.4 Events such as a fire, airborne activity or chemical contamination could render the Main Control Room uninhabitable. The Remote Shutdown System Panels in conjunction with other remotely operable systems/equipment are used to shutdown the reactor and bring it to a cold shutdown condition.

1.5 It is assumed that the Remote Shutdown Panels and associated ECCS/ESF systems are lined up in standby per SOP-0027, Remote Shutdown System.

1.6 The following outlines the responsibilities of the Operations Staff in the performance of this AOP. This listing is provided for guidance only and does not limit the OSM/CRS from assigning personnel as deemed necessary.

1.6.1. The OSM implements the EIPs and assumes the position of Emergency Director. The OSM may report to the Remote Shutdown Panel to perform an initial assessment, but in all cases must report to the TSC to implement the EIPs within the time frame specified. Once relieved, the OSM may return to the Remote Shutdown Panel and assume the position of advisor to the CRS during the use of EOPs and this AOP.

1.6.2. The CRS assumes the position of Director/Coordinator during use of EOPs and this AOP.

1.6.3. The ATC Operator implements AOP-0031 and EOPs as directed by CRS.

1.6.4. The Unit Operator implements AOP-0031 and EOPs as directed by CRS.

1.6.5. The Control Bldg Operator performs duties as directed by OSM; reports to OSC.

1.6.6. The Reactor Bldg Operator implements AOP-0031 and EOPs as directed by CRS.

1.6.7. The Turbine Bldg Operator performs duties as directed by OSM.

1.6.8. The Outside Operator performs duties as directed by OSM; reports to OSC.

1.6.9. The Aux Control Room Operator implements AOP-0031 and EOPs as directed by CRS.

1.6.10. The Radwaste Operator performs duties as directed by OSM.

1.6.11. The STA reports to the Div 1 Remote Shutdown Panel to provide oversight/support.

1.6.12. The MCR communicator reports to the TSC until relieved, at which time, the OSM assigns his duties.

1.7 The ATC Operator Actions, especially when a Main Control Room fire is in progress, are necessary to first ensure the reactor is shut down. Then to man the Div I Remote Shutdown Panel and transfer switches such that RPV parameters, SRV and RCIC controls are available for use to monitor and control RPV pressure and level within 15 minutes of scramming the reactor. It is expected that RCIC will be the initial source of makeup to the RPV once this transfer has been completed.

1.8 If a Main Control Room fire is in progress, the Aux Control Room Operator will isolate all 10 Condensate Demins within 3 minutes of the reactor being scrammed as directed by the CRS. This should trip and prevent restart of all Reactor Feed Pumps due to possible hot shorts. (Ref 6.18) 1.9 If a Main Control Room fire is in progress, the UO Operator will locally initiate a Reactor Scram and a full NSSSS Isolation at the RPS MG sets and EPA breakers at CB 116 within 5 minutes of scramming the reactor. Then the UO will deenergize ENB-PNL02A and ENB-PNL02B within 10 minutes of scramming the reactor. This should close any SRVs which may have opened due to possible hot shorts. Once this has been completed further actions will align the plant for Remote Operation systems necessary for plant cooldown including electrical power, HVAC, and cooling systems.

1.10 If a Main Control Room fire is in progress, the RB Operator will first transfer power, for E51-MOVF063 from Div II to Div I Alternate Power in less than 10 minutes of scramming the reactor. Once this has been completed further actions will align the plant for Remote Operation systems necessary for plant cooldown including electrical power, HVAC, and cooling systems.

Examination Outline Cross-Reference Level RO 259002 Reactor Water Level Control Tier # 2 Group # 1 Knowledge of the effect that a loss or K/A # K6.02 malfunction of the following will have on the Rating 3.3 REACTOR WATER LEVEL CONTROL SYSTEM: K6.02 A.C. power Question 37 The reactor is operating at 55 % rated thermal power.

Three Reactor Feed Pumps are operating The 480 volt breaker NHS-MCC1CA for FWL-P1A Reactor Feed Pump Main Oil Pump trips on overcurrent.

Based on the provided information what is the expected response of the Reactor Vessel Level?

A. RPV level will initially lower and then be restored to programmed level B. RPV Level will initially lower and then stabilize at a new lower level C. RPV Level will remain constant D. RPV Level will lower to the point of an automatic reactor scram Answer: C Explanation:

A. Incorrect, The Aux Lube oil pump will auto start at a sensed pressure of 5 psig and maintain the necessary oil pressure to maintain the RFP operating B. Incorrect, The Aux Lube oil pump will auto start at a sensed pressure of 5 psig and maintain the necessary oil pressure to maintain the RFP operating C. Correct D. Incorrect, The Aux Lube oil pump will auto start at a sensed pressure of 5 psig and maintain the necessary oil pressure to maintain the RFP operating Technical

References:

R-STM-0107 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content: 55.41(b)(7) d) Automatic Functions and Interlocks (1) Main Oil Pumps (FWL-P l)

These pumps stop if the control switch is in AUTO and the associated feed pump is stopped for 23 minutes.

(2) Auxiliary (AC) Oil Pumps (FWL-P2)

The respective pump automatic logic is as follows:

• The reactor feed pump breaker must be closed (i.e., RFP pump motor overloads reset).

RFP pump breaker is required to be closed to initially arm the auto start logic.

• FWL-P2 control switch in AUTO This "arms" the auto start feature on low lube oil pressure (:.:; 5 psig).

Once armed, the RFP pump breaker can be opened and the circuit will remain armed.

(a) On low lube oil pressure, :.:; 5 psig as sensed by PS-11, the pump auto starts and continues to run regardless of lube oil header pressure until the pump is turned off by placing its control switch in the STOP position.

Examination Outline Cross-Reference Level RO 400000 Component Cooling Water Tier # 2 Group # 1 Knowledge of electrical power supplies to the K/A # K2.01 flowing: CCW pumps Rating 2.9 Question 38 What are the power supplies to the RPCCW (CCP) pumps?

A. CCP-P1A NNS-SWG1A CCP-P1C NNS-SWG1C B. CCP-P1A NNS-SWG1A CCP-P1C NNS-SWG1A C. CCP-P1A NJS-SWG1A CCP-P1C NJS-SWG1C D. CCP-P1A NJS-SWG1A CCP-P1C NJS-SWG1A Answer: D Explanation:

A is wrong -these are power supplies for TPCCW pumps B is wrong - these are power supplies for TPCCW pumps (one letter switched)

C is wrong - these are the correct switchgears however the C pump is also off of the A bus so this is incorrect.

D is correct -Both pumps are off of the NJS switchgear and on both are on the A bus.

Technical

References:

R-STM-0115, REV 6, REACTOR PLANT COMPONENT COOLING WATER (CCP) SYSTEM References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0115 Obj E5 Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO 239002 SRVs Tier # 2 Group # 1 Knowledge of the operational implications of K/A # K5.05 the following concepts as they apply to Rating 2.6 RELIEF/SAFETY VALVES: Discharge line quencher operation Question 39 Per EOP basis, which one of the following is the highest suppression pool 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 (SRV Tail Pipe Level Limit)?

A. 15 feet 5 inches B. 19 feet 6 inches C. 20 feet D. 21 feet 3 inches Answer: D Explanation:

A is wrong because when suppression pool level decreases to two feet above the top of the Mark III horizontal vents, any further drop in water level could result in direct exposure of the drywell atmosphere to the containment airspace thus compromising the pressure suppression function of the containment. Suppression pool level should therefore be maintained above this elevation. Per EOP-2 Step SPL-9, this level is 15 feet 5 inches.

B is wrong because per Tech Spec 3.6.2, 19 feet 6 inches is the minimum operating suppression pool level in modes 1, 2, and 3.

C is wrong because per Tech Spec 3.6.2, 20 feet is the maximum operating suppression pool level in modes 1, 2, and 3.

D is correct because per EOP basis, the STPLL is 21 feet 3 inches. Since SRV operation with suppression npool level above the STPLL could lead to containment failure, the RPV is not permitted to remain at pressure if suppression pool level will exceed this level.

The SRV Tail Pipe Level Limit (STPLL) is the lesser of:

• The Maximum Pressure Suppression Primary Containment Water Level.

• The highest suppression pool 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.

Technical

References:

EPSTG*0002, Page B-8-25 Rev 17 References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0109, Objective 3 f.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

I changed the wording to the stem and removed the 17 feet distractor. I thought we needed another high level distractor. I also included the two references. The figure with the quencher height is attached on the last page. - GJK 6-16-2016

EOP-2 Step (SPL-7, SPL-8)

Discussion The SRV Tail Pipe Level Limit (STPLL) is the lesser of:

• The Maximum Pressure Suppression Primary Containment Water Level.

• The highest suppression pool 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.

SRV operation with suppression pool water level above the STPLL could damage the SRV discharge lines. This, in turn, could lead to containment failure from direct pressurization and damage to equipment inside the containment (ECCS piping, RPV water level instrument runs, etc.) from pipe-whip and jet-impingement loads. Refer to Appendix A for a detailed discussion of the STPLL.

At River Bend, the STPLL is 21 ft. 3 in. and is constant throughout the entire range of expected RPV pressures. Since SRV operation with suppression pool level above the STPLL could lead to containment failure, the RPV is not permitted to remain at pressure if suppression pool level will exceed this level.

Consistent with the definition of restore, emergency depressurization is not required until it has been determined that actions to restore suppression pool level below the STPLL will not be effective.

Examination Outline Cross-Reference Level RO 295028 High Drywell Temperature Tier # 1 Group # 1 Ability to operate and/or monitor the following K/A # EA1.02 as they apply to high drywell temperature: Rating 3.9 Drywell ventilation system Question 40

• The plant is operating at 100% power

• The control switches for drywell unit coolers DRS-UC1A through -UC1D are in RUN

• Temperature controllers DRS-H/A52A through -H/A52D are in AUTO

• Plant Engineering and responsible System Engineers have been briefed on current plant configuration

• On H13-P808, CMS-TR41A and -TR41B read 1100F and have been stable for days a) Per procedure SOP-0060, what are the minimum permissible number of drywell unit coolers that can be running at this time?

b) If CMS-TR41A and -TR41B change to 1500F, what is the maximum number of drywell unit coolers that can be running?

a) b)

A. 4 5 B. 4 6 C. 5 5 D. 5 6 Answer: B Explanation:

A is wrong. The answer for part a) is correct. However, since the plant is in an emergency condition, six drywell cooling units can be running. This is plausible if the applicant believes the prohibition against running all six units to prevent the airflow stall condition applies in this situation. However, this applies during normal operations.

B is correct. Procedure SOP-0060, Drywell Cooling (SYS #404), Section 4.1 contains a Caution and a Note which provides stipulations on how many drywell cooling units can be run in different situations. Per the Notes, it is permissible to run four drywell unit coolers, provided that the average drywell temperature is less than or equal to 1450F, and Plant Engineering and the System Engineer are notified. Therefore, 4 for a) is correct. For b), this involves a condition where EOP-2 would be entered with Drywell temperature greater than 1450F. Since this constitutes an emergency condition, all six of the drywell cooling units can be running, per the Caution in the procedure.

C is wrong. Part a) is plausible if the conditions for allowing only 4 units to run had not been met, as stated in the procedure Notes. However, they are met. For part b), this is the maximum number for normal operations. Emergency operations, due to EOP-2 entry, are in effect.

D is wrong due to the answer for part a). See the explanation for answer C.

Technical

References:

SOP-0060, Drywell Cooling (SYS #404), Revision 10 EOP-2, Primary Containment Control, Revision 16 System Training Manual R-STM-0057, Primary Containment and Auxiliaries, Revision 5 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H4 10CFR Part 55 Content: 55.41(b)(7) 55.43

Examination Outline Cross-Reference Level RO 268000 Radwaste Tier # 2 Group # 2 Ability to predict and/or monitor changes in K/A # A1.01 parameters associated with operating the Rating 2.7 RADWASTE controls including: Radiation level Question 41

• A liquid effluent discharge of a Recovery Sample Tank is in progress.

• An alarming condition caused by a High Alarm as measured by RMS-RE107, Liquid Radwaste Radiation Monitor, comes in at the Auxiliary Control Room (ACR).

In the Main Control Room, this condition can be confirmed by viewing a channel display color the associated radiation monitor of a) on panel b) .

A. a) Red b) DSPL230 B. a) Red b) P878 C. a) Yellow b) P878 D. a) Yellow b) DSPL230 Answer: A Explanation:

A is correct. a) When a liquid effluent discharge is prepared for, Alert and Alarm levels are established for RMS-RE107. These values are input into the Digital Radiation Monitoring System (DRMS). When the channel display for the radiation monitor achieves its Alarm level, it is Red in color. See R-STM-0511, Section C.3.a). CSP-0110, Definitions 3.27 and 3.28, confirm that Alert Alarm and High Alarm set points are calculated. To match the DRMS scheme, these would correspond to its Alarm and Alert points.

For b), this radiation monitor, a non safety-related monitor, only has monitoring available via the DRMS in the Main Control Room. This is displayed in panel DSPL230. This interface is referred to in SOP-0113 and SOP-0086.

B is wrong. The channel display color is correct. However, panel P878 provides MCR safety-related radiation monitor indication.

C is wrong. The Alert color of Yellow is assigned to a different value than that assigned to the ACR alarm and the Alarm level set in DRMS. The panel doesnt provide indication for the radiation monitor in question.

D is correct. The panel is correct, but the display color is incorrect.

Technical

References:

R-STM-0603, Radwaste Systems, Revision 7

R-STM-0511, Radiation Monitoring, Revision 15 SOP-0113, Liquid Radwaste Processing/Recovery Sample Tank System (SYS #603),

Revision 24 SOP-0086, Digital Radiation Monitoring System (SYS #511), Revision 16 ADM-0054, Radioactive Liquid Effluent Batch Discharge, Revision 6A ARP-RMS-DSPL230, DRMS Computer CRT (RMS-DSPL230) Alarm Response, Revision 9 CSP-0110, Radioactive Liquid Effluent Batch Discharge, Revision 19 ARP-LWS-PNL187-4, LWS-PNL18704 Alarm Response, Revision 303 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(5)

Examination Outline Cross-Reference Level RO Equipment Control Tier # 3 Group #

Ability to manipulate console controls as K/A # 2.2.2 requird to operate the facility between Rating 4.6 shutdown and power levels Question 42 When operating a motor-operated throttle valve from the control room, the hand switch should be held in the ___(1)______ because ___(2)______.

A. 1) closed position for at least 5 seconds after the red light extinguishes

2) the red light is set to extinguish before the valve reaches its closed seat B. 1) open position for at least 3 seconds after the green light extinguishes

2) the green light is set to extinguish before the valve starts to come off its open seat C. 1) closed position for at least 3 seconds after the red light extinguishes

2) the red light is set to extinguish before the valve reaches its closed seat D. 1) open position for at least 5 seconds after the green light extinguishes

2) the green light is set to extinguish before the valve starts to come off its open seat Answer: A Explanation:

A is correct per OSP-0022, page 11.

B is incorrect because the time is wrong (it is 5 sec not 3 sec) and also it is for the red light and its closed seat not the green light and the open seat.

C is incorrect because the time is wrong (it is 5 sec not 3 sec) the second part is correct.

D is incorrect because of the second part, it is set to the red light and the closed seat but the time of 5 sec is correct.

Technical

References:

OSP-0022, rev 86, page 11.

References to be provided to applicants during exam: None.

Learning Objective: RB to provide these (OSP Los not sent)

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)10

Examination Outline Cross-Reference Level RO 219000 RHR/LPCI: Torus/Suppression Pool Tier # 2 Cooling Mode Group # 2 K/A # K2.01 Knowledge of electrical power supplies to the Rating 2.5 following: Valves Question 43 Which of the following is the correct power supply for the RHR B & D Heat Exchanger Shell Bypass Valve E12-MOV048B?

A. EHS-MCC2E B. EHS-MCC2F C. EHS-MCC2G D. EHS-MCC2H Proposed Answer: B Explanation (Optional):

A. INCORRECT. EHS-MCC2E is the power supply for E12-MOV048A.

B. CORRECT. EHS-MCC2F is the power supply for E12-MOV048B.

C. INCORRECT. EHS-MCC2G is the power supply for Division I RHR valves.

D. INCORRECT. EHS-MCC2H is the power supply for Division II RHR valves, but not the heat exchanger bypass valve.

Valve is important because after recovering reactor water level, it may be necessary to place at least one of the RHR loops in the suppression pool cooling mode of operation.

Ten minutes after the initiation signal is received, the open signal to the heat exchanger bypass valves MOVF048A(B) is removed, allowing them to be throttled back or closed as required.

Technical Reference(s):

STM-0204, Residual Heat Removal System (RHR), Revision 12 Proposed references to be provided to applicants during examination: None Learning Objective: RLP-STM-0204, NLO Objective F (f)

Question Source: New Question History: N/A Question Cognitive Level: Memory or Fundamental Knowledge F4 10 CFR Part 55 Content: 55.41.7 Comments:

Examination Outline Cross-Reference Level RO 203000 RHR/LPCI: Injection Mode Tier # 2 Group # 1 Knowledge of the effect that a loss or K/A # K3.04 malfunction of the RHR/LPCI: injection mode Rating 4.6 (plant specific) will have on following: K3.04 Adequate core cooling Question 44 During an emergency the SRO (CRS) is making the determination that an emergency depressurization is required. The SRO (CRS) asks for a status update and you give the following information:

RHR PUMP A IN MAN OVERRIDE Annunciator lit RHR B has an amber light lit for its associated injection valve RHR C has a red light and a white light lit for its associated pump control switch E22-S0011G1C bus fault No other equipment is available for use and LOCA signals are still present.

Adequate core cooling is assured by A. RHR A, and RHR B only B. RHR B and RHR C only C. RHR C only D. HPCS and LPCS only Answer: C Explanation:

A is wrong because RHR A pump has a manual override in. The pump would need to be reset by pushing the reset pushbutton and the auto initiation signal would need to be clear.

B is wrong because the amber light indicates the valve has been manually overridden similar to RHR A.

C is correct because RHR C has proper indications for a pump in an ED situation and is ready to refill the core which would lead to core submergence and therefore adequate core cooling.

D is wrong because HPCS bus has a fault making it inoperable.

May need to add a faulted condition for LPCS.

Technical

References:

R-STM-0203, High Pressure Core Spray, Rev 8 R-STM-0204, Residual Heat Removal System. Rev 12 R-STM-0205, Low Pressure Core Spray, Rev 6 References to be provided to applicants during exam: None.

Learning Objective: List the automatic functions and interlocks for each of the following Residual Heat Removal System components:

LPCI Injection Isolation Valves, E12-MOVF042 Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(8)

Examination Outline Cross-Reference Level RO 264000 EDGs Tier # 2 Group # 1 Knowledge of the effect that a loss or K/A # K3.01 malfunction of the EMERGENCY Rating 4.2 GENERATORS (DIESEL/JET) will have on following: Emergency core cooling systems Question 45 The station has experienced a Loss of Coolant Accident concurrent with a Loss of Off-site Power. In this situation, a ____(1)____ on the Division 1 Emergency Diesel Generator would cause the loss of ____(2)___ Division 1 ECCS pumps needed for the LOCA conditions.

A. (1) Ground fault (2) LPCS and RHR A only B. (1) Ground fault (2) RHR A and RHR C only C. (1) Sustained lube oil pressure of 20 psig (2) LPCS and RHR A only D. (1) Sustained lube oil pressure of 20 psig (2) RHR A and RHR C only Answer: A Explanation:

A is correct because LO pressure trips are disabled for LOCA events while a ground fault is not. The second part is knowing what the division 1 EDG powers and the answer is LPCS and RHR A pump only.

B is incorrect because the second part is wrong. RHR pump C is off of the division 2 bus/EDG. The first part is correct.

C is incorrect because LO pressure trips are disabled for LOCA events and the second part is correct.

D is incorrect because LO pressure trips are disabled for LOCA events and the pumps are incorrect because RHR pump C is off of the division 2 bus/EDG.

Technical

References:

STM-309S, revision 13.

References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-309S, 5A, 6D.

Question Source: Bank #

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New X Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO 295030 Low Suppression Pool Wtr Lvl Tier # 1 Group # 1 Ability to determine and/or interpret the K/A # EA2.03 following as they apply to low suppression Rating 3.7 pool water level: EA2.03 Reactor pressure Question 46 An ATWS has occurred following an inadvertent MSIV isolation.

Which of the following reactor pressure bands are appropriate for the given suppression pool temperatures and levels?

Reactor Pressure Band Supp Pool Level Supp Pool Temp A. 500 - 700 psig 15 4 130 oF B. 800 - 1090 psig 16 11 128 oF C. 800 - 1090 psig 19 5 140 oF D. 500 - 700 psig 21 3 150 oF Answer: A Explanation:

A is correct because it is the only choice that allows full use of the pressure band without exceeding the HCTL curve, therefore it is correct.

B is wrong because its band falls within the unsafe region (exceeds the HCTL curve)

C is wrong because its band falls within the unsafe region (exceeds the HCTL curve)

D is correct because its band falls within the unsafe region (exceeds the HCTL curve)

Technical

References:

EOP-0001 Figure 2 HCTL curve, revision 27.

References to be provided to applicants during exam: EOP-0001 Figure 2 Learning Objective: RLP-HLO-517 Obj 2 Question Source: Bank # X (2010 NRC)

(note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)10

Examination Outline Cross-Reference Level RO 295023 Refueling Accidents Tier # 1 Group # 1 Ability to determine and/or interpret the K/A # AA2.02 following as they apply to refueling accidents: Rating 3.4 Fuel Pool Level Question 47 During movement of irradiated fuel in the Spent Fuel Storage Pool, Storage Pool level starts to drop.

Which of the following is the minimum Technical Specification level that is required to be maintained over the irradiated fuel assemblies in the Spent Fuel Storage Pool?

A. 21 B. 22 C. 23 D. 24 Proposed Answer: C Explanation (Optional):

A. INCORRECT. According to TS 3.7.6, the fuel pool water level shall be 23 ft over the top of irradiated fuel assemblies.

B. INCORRECT. According to TS 3.7.6, the fuel pool water level shall be 23 ft over the top of irradiated fuel assemblies.

C. CORRECT. According to TS 3.7.6, the fuel pool water level shall be 23 ft over the top of irradiated fuel assemblies.

D. INCORRECT. According to TS 3.7.6, the fuel pool water level shall be 23 ft over the top of irradiated fuel assemblies.

Technical Reference(s):

Technical Specification 3.7.6, Amendment 81 Proposed references to be provided to applicants during examination: None Learning Objective: RLP-HLO-0416, Objective 2 Question Source: Bank Question History: Last NRC Exam Columbia 2011-04 Exam, Question 10 Question Cognitive Level: Memory or Fundamental Knowledge F2 10 CFR Part 55 Content: 55.41.10 Comments:

Columbia 2011-04 Exam, Question 10

Examination Outline Cross-Reference Level RO 201001 CRD Hydraulic Tier # 2 Group # 2 Ability to interpret control room indications to K/A # 2.2.44 verify the status and operation of a system, Rating 4.2 and understand how operator actions and directives affect plant and system conditions.

Question 48 While performing friction testing on a control rod the control rod that is selected moves an additional half-step than demanded with -- (double blank) being indicated as the control rod position. The Control Rod Drift annunciator would be (A) and the control rod would be moved (B).

A. lit; in to 00 B. lit; out to original position C. extinguished; in to 00 D. extinguished; out to original position Answer: D Explanation:

A is wrong because the annunciator would not be lit but if it was it would be required to be inserted to 00.

B is wrong because the annunciator would not be lit.

C is wrong because the operator would not be required to move the rod to 00 since the rod is not drifting.

D is correct because with a normal insert command the control rod drift annunciator would not illuminate. Due to the rod not going in more than one additional step the operator would be allowed to pull the rod bank out to the original step per procedure guidance.

Technical

References:

AOP-61, Control Rod(s) Mispositioned/Malfunction, Rev 8 STP-052-0102, Partially Withdrawn Control Rod Insertion Operability Check, Rev 9 References to be provided to applicants during exam: None.

Learning Objective: Describe the controls / indications / interlocks for the following:

a) Control Rod Drift Describe the function(s) of the Control Rod Drive Mechanism during:

a) Control Rod Withdraw b) Control Rod Insertion Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(6)

Examination Outline Cross-Reference Level RO Tier # 3 Group #

Knowledge of shift or short-term relief turnover K/A # 2.1.3 practices. Rating 3.7 Question 49 You are the oncoming ATC. Which of the following MUST be completed PRIOR to assuming the shift?

A. Review LCO/Tracking LCO Logs B. Review Night/Standing Orders C. Log Reporting Equipment Status D. Annunciator Status Review Proposed Answer: B Explanation (Optional):

A. INCORRECT. This is to be completed as early in the shift as possible.

B. CORRECT. According to Attachment 3, this must be completed prior to assuming the shift.

C. INCORRECT. This is to be completed as early in the shift as possible.

D. INCORRECT. This is to be completed as early in the shift as possible.

Technical Reference(s):

OSP-002, Shift Relief and Turnover, Rev 47 Proposed references to be provided to applicants during examination: None Learning Objective: RLP-HLO-0206, Objective D Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge F2 10 CFR Part 55 Content: 55.41.10 Comments:

K/A Statement: 2.1.3 Knowledge of shift or short-term relief turnover practices.

Examination Outline Cross-Reference Level RO 261000 SGTS Tier # 2 Group # 1 Ability to monitor automatic operations of the K/A # A3.04 STANDBY GAS TREATMENT SYSTEM Rating 3.0 including: System Temperature Question 50 A plant event occurs that causes the automatic start of both trains of Standby Gas Treatment System (GTS).

A control room operator can monitor the charcoal bed inlet temperatures at __(a)__.

Charcoal bed temperature exceeding __(b)__ would cause the annunciator SGT FILTER TRAIN FLT1A(B) CHARCOAL TEMPERATURE HIGH to alarm.

A. a) panel H13-P863 b) 233.50F B. a) panel H13-P863 b) 2500F C. a) process computer b) 233.50F D. a) process computer b) 2500F Answer: A Explanation:

A is correct. Per R-STM-0257, Page 10 of 28, GTS-RTD27A(B) provides temperature indication of the charcoal bed inlet temperature on panel H13-P863. Table 4 of R-STM-0257 indicates that the selected annunciator comes in when charcoal bed temperature exceeds 233.50F.

B is wrong. Part a) is correct, but b) is incorrect. 2500F is the temperature that the SGT FILTER TRAIN GTS-FLT1A(B) INOPERATIVE alarm comes in via GTS-TS5A. See R-STM-0257, Table 4.

C is wrong. GTS-RTD14A(B), which does provide indication on the process computer, is an indication of charcoal bed outlet temperature.

D is wrong. See the Explanation for answers B and C.

Technical

References:

R-STM-0257, Standby Gas Treatment System, Revision 5 ARP-863-73, P863-73 Alarm Response, Revision 8 SOP-0043, Standby Gas Treatment System (SYS #257), Revision 17 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 295024 High Drywell Pressure Tier # 1 Group # 1 Ability to interpret control room indications to K/A # 2.2.44 verify status and operation of a system, and Rating 4.2 understand how operator actions and directives affect plant and system conditions Question 51 Complete the following statements regarding High Drywell Pressure RPS trip units.

1. Upon a loss of 24 VDC power to an entire RPS trip channel, the High Drywell Pressure Trip unit _______ trip to provide a SCRAM input.

2. Operators _____ able to bypass a High Drywell Pressure Trip unit.

A. will not; are B. will not; are not C. will; are D. will; are not Answer:

Explanation:

A is wrong because the high drywell pressure trip units are deenergize to actuate and are powered from 24 VDC power, not 120 VAC power. Plausible if applicant believes the trip is energized to actuate or powered from 120 VDC.

B is wrong because see A C is wrong because no bypass option is provided for the high drywell scram. Plausible if applicants believe it is one of the scrams that can be bypassed.

D is correct because the High drywell trip units are powered from 24 VDC and have no bypass option.

Technical

References:

R-STM-0508, Reactor Protection System References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO Equipment Control Tier # 3 Group #

Knowledge of pre- and post-maintenance K/A # 2.2.21 operability requirements. Rating 2.9 Question 52 In accordance with Admin Procedure ADM-0015, the surveillance procedure performers are required to ___(1)_____ when the ___(2)_____.

A. 1) use equivalent M &TE to confirm readings

2) obtained values are out of tolerance but do not exceed the TS tolerance B. 1) use different M &TE to confirm readings

2) obtained values are out of tolerance but do not exceed the TS tolerance C. 1) use equivalent M &TE to confirm readings

2) obtained values are greater than twice the tolerance D. 1) use different M &TE to confirm readings

2) obtained values are greater than twice the tolerance Answer: C Explanation:

A is incorrect-although the first part is correct it is only when obtained values are out of tolerance by more than twice the value per section 4.8.9, page 9 of ADM-0015.

B is incorrect- first part is incorrect must use equivalent M & TE and second part is also incorrect.

C is correct- per section 4.8.9, page 9 of ADM-0015, must use equivalent M&TE for tolerances that exceed TS allowed tolerance or if twice the allowed tolerance to verify the values for the ST.

D is incorrect because the first part is incorrect- see above for explanation.

Technical

References:

ADM-0015, rev 38, page 9, section 4.8.9.

References to be provided to applicants during exam: None.

Learning Objective: HLO-0221 Obj 1a and Obj 9 Question Source: Bank #

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New X Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis F3 10CFR Part 55 Content: 55.41(b)(10)

Examination Outline Cross-Reference Level RO 201005 RCIS Tier # 2 Group # 2 Knowledge of the operational implications of K/A # K5.09 the following concepts as they apply to ROD Rating 3.5 CONTROL AND INFORMATION SYSTEM (RCIS): (CFR: 41.5 / 45.3)

K5.09 High power setpoints BWR-6 Question 53 Which of the following describes the operational implications of maintaining control rods within designated withdrawal distances during control rod movements above the high power setpoint?

A. Establish a 2 notch limit to mitigate the consequences of a control rod drop accident by limiting the amount and rate of reactivity increase.

B. Establish a 4 notch limit to mitigate the consequences of a control rod drop accident by limiting the amount and rate of reactivity increase.

C. Establish a 2 notch limit to provide protection for a control rod withdrawal error event to preclude a MCPR safety limit violation.

D. Establish a 4 notch limit to provide protection for a control rod withdrawal error event to preclude a MCPR safety limit violation.

Answer:

Explanation:

A. incorrect - the rod pattern controller of RC&IS is designed to mitigate the consequences of a control rod drop accident B. incorrect - the rod pattern controller of RC&IS is designed to mitigate the consequences of a control rod drop accident C. Correct D. incorrect - the 4-notch limit is enforced between the LPSP and the HPSP Technical

References:

Tech Spec Bases 3.3.2.1 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(6)

B 3.3.2.1 Control Rod Block Instrumentation BASES BACKGROUND Control rods provide the primary means for control of reactivity changes. Control rod block instrumentation includes channel sensors, logic circuitry, switches, and relays that are designed to ensure that specified fuel design limits are not exceeded for postulated transients and accidents. During high power operation, the rod withdrawal limiter (RWL) provides protection for control rod withdrawal error events. During low power operations, control rod blocks from the rod pattern controller (RPC) enforce specific control rod sequences designed to mitigate the consequences of the control rod drop accident (CRDA). During shutdown conditions, control rod blocks from the Reactor Mode Switch-Shutdown Position ensure that all control rods remain inserted to prevent inadvertent criticalities.

The purpose of the RWL is to limit control rod withdrawal to preclude a MCPR Safety Limit (SL) violation. The RWL supplies a trip signal to the Rod Control and Information System (RCIS) to appropriately inhibit control rod withdrawal during power operation equal to or greater than the low power setpoint (LPSP).

Examination Outline Cross-Reference Level RO Radiation Control Tier # 3 Group #

Knowledge of radiological safety principles K/A # 2.3.12 pertaining to licensed operator duties, such as Rating 3.2 containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Question 54 In accordance with TS 5.7, High Radiation Area:

The lowest area radiation level where locked or continuously guarded doors are required to prevent unauthorized entry is _____.

Entry into a locked high radiation area is permitted under an approved RWP that shall specify the dose rate levels in the immediate work areas and ____________

individuals in those areas.

A. 1000 mrem/hr; periodic RP surveillance of B. 500 mrem/hr; periodic RP surveillance of C. 500 mrem/hr; maximum allowable stay times for D. 1000 mrem/hr; maximum allowable stay times for Answer:

Explanation:

A is wrong because periodic surveillance is incorrect. See B.

B is wrong because 1000 mrem/hr is the lowest radiation level that requires locked/guarded doors. Plausible if applicant confuses 500 rad/hr limit for VHRA. Continuous surveillance is required in lieu of max stay times. Plausible if applicant considers periodic surveillance to be adequate.

C is wrong because 500 mrem/hr is incorrect. See B.

D is correct because 1000 mrem is the lowest level requiring. Max stay times designated in the RWP meets the requirement.

Technical

References:

TS 5.7.2 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(12)

Examination Outline Cross-Reference Level RO 223001 Primary CTMT and Aux. Tier # 2 Group # 2 Knowledge of the effect that a loss or K/A # K6.02 malfunction of the following will have on the Rating 3.5 PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES: Containment cooling: Mark-III Question 55 During surveillance testing I&C personnel inadvertently isolated all containment chilled water supply and return isolation valves. What is the MAXIMUM containment or drywell temperature that would require both the mode switch to be placed in shutdown and an emergency depressurization?

A. 110°F B. 145°F C. 165°F D. 185°F Answer: D Explanation:

185°F containment temperature is the correct answer. 110°F is the suppression pool temperature requiring the mode switch to be placed in shutdown but would not require an ED. 145°F drywell temperature would require a shutdown but not an ED. 165°F is the maximum post LOCA long term bulk air temp. This is more of a design temperature/limit for containment. 330°F drywell temperature would also require a shutdown and an ED.

Technical

References:

R-STM-0057, PRIMARY CONTAINMENT AND AUXILIARIES, Revision 5, p. 6-12 References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0057, REV. 7, ATTACHMENT 3, Objective D Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(9)

Examination Outline Cross-Reference Level RO Tier # 3 Knowledge of the parameters and logic used Group #

to assess the status of safety functions, such K/A # 2.4.21 as reactivity control, core cooling and heat Rating 4.0 removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Question 56 River Bend has entered EOP 4, RPV Flooding, and are injecting via the suppression pool.

Which of the following are NOT indications that the RPV has been flooded and main steam lines are full, according to EOP-4, RPV Flooding?

A. Lowering RPV pressure B. Steam line flows C. Actuation of SRV tailpipe acoustic monitors D. Stable suppression pool level Proposed Answer: A Explanation (Optional):

A. CORRECT. According to the table in EOP-4, rising RPV pressure is an indication of a flooded RPV.

B. INCORRECT. Plausible if one believes that the steam line flows indicate steam flow.

C. INCORRECT. Plausible if one believes that the acoustic monitors are measuring the sound of SRV steam flow.

D. INCORRECT. Plausible if one believes that suppression pool level should be decreasing due to injection from the suppression pool.

Technical Reference(s):

EOP-4, RPV Flooding, Revision 15 Proposed references to be provided to applicants during examination: None Learning Objective: R-LPOPS-HLO-511, Objective D Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge F2 10 CFR Part 55 Content: 55.41.7 Comments:

Examination Outline Cross-Reference Level RO 290002 Reactor Vessel Internals Tier # 2 Group # 2 Ability to explain and apply system limits and K/A # 2.1.32 precautions. Rating 3.8 Question 57 Which of the following precautions regarding Single Loop Operation is correct, and the reason for the precautions?

A. A multiplier of 0.83 is applied to the APLHGR, to ensure that the fuel cladding 1% plastic strain is not exceeded.

B. A multiplier of 0.83 is applied to the APLHGR, to ensure that 99.9% of the fuel rods avoid boiling transition if the limit is not violated.

C. A multiplier of 0.83 is applied to the MCPR, to ensure that the fuel cladding 1% plastic strain is not exceeded.

D. A multiplier of 0.83 is applied to the MCPR, to ensure that 99.9% of the fuel rods avoid boiling transition if the limit is not violated.

Proposed Answer: A Explanation (Optional):

A. CORRECT. When in Single Loop Operation, and a multiplier of 0.83 is applied to the APLHGR. The 1% plastic strain is the limit for APLHGR.

B. INCORRECT. First part is correct. Second part is incorrect but plausible if one confuses the safety limits for APLHGR and MCPR.

C. INCORRECT. During Single Loop Operation, MCPR is increased by 0.02. Part 1 is plausible if one confuses MCPR and LHGR, which does have a 0.83 multiplier. Part 2 is correct for APLHGR.

D. INCORRECT. During Single Loop Operation, MCPR is increased by 0.02. Part 1 is plausible if one confuses MCPR and LHGR, which does have a 0.83 multiplier. Part 2 is correct for MCPR, thus plausible if one confuses APLHGR and MCPR.

Technical Reference(s): GOP-004, Single Loop Operation, Revision 22, Core Operating Limits Report, Cycle 19, Revision 1 Proposed references to be provided to applicants during examination: None Learning Objective:

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge F4 10 CFR Part 55 Content: 55.41.10

Comments:

Examination Outline Cross-Reference Level RO 300000 Instrument Air Tier # 2 Group # 1 Knowledge of the operational implications of K/A # K5.01 the following concepts as they apply to the Rating 2.5 INSTRUMENT AIR SYSTEM: Air compressors Question 58 The instrument air compressors are all available. Air compressor IAS-C2C is currently the only one operating to maintain system pressure and the local SEQUENCE CONTROL switch is in Position 3 (C-A-B).

A relay failure in the control circuit for compressor IAS-C2C causes it to shutdown.

With no change in Instrument Air System usage, which of the following describes the effect of the compressor shutdown on the Instrument Air System?

A. IAS-C2A operates alone, maintaining a lower header pressure.

B. IAS-C2A operates alone, maintaining the same header pressure.

C. Both IAS-C2A and C2B are operating, maintaining the same header pressure.

D. Both IAS-C2A and C2B are operating, maintaining a lower header pressure.

Answer: A Explanation:

A is Correct-IAS-C2A is the second compressor to sequence on in Position 3, so it will start as pressure lowers to the mid range setpoint (115.5 psig) which is lower than the pressure setpoint for the start of IAS-C2C (118.5 psig).

B is wrong because it will not maintain the same header pressure.

C is wrong because system demand is unchanged, so there is no need for two compressors to run. Additionally, system pressure will not be maintained the same as before the shutdown (118.5 psig) because IAS-C2A is being controlled by the mid range pressure switches (115.5 psig) in Position 3.

D is wrong because system demand unchanged, so there is no need for two compressors to run.

Technical

References:

R-STM-0121 Rev 16 Pg 14 of 69 References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0121 Obj 3a Question Source: Bank # X (note changes; attach parent) Modified Bank #

New

Question History: Last NRC Exam 2012 Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)4

Examination Outline Cross-Reference Level RO 295015 Incomplete SCRAM / 1 Tier # 1 Group # 2 Ability to identify post-accident K/A # 2.4.3 instrumentation. Rating 3.7 Question 59 Which of the following neutron flux instruments are the least sensitive to failure-to-scram neutron flux oscillations which can lead to localized fuel damage?

A. LPRMs B. IRMs C. SRMs D. APRMs Answer:

Explanation:

A is wrong because... Local neutron flux instrumentation (LPRMs, IRMs, SRMs) will detect failure-to-scram neutron flux oscillations more accurately than APRMs.

B is wrong because See A.

C is wrong because See A.

D is correct because APRMs will be the least sensitive to oscillations due to its associated averaging circuitry. Plausible as the Period-Based Detection System (PBDS) cards are mounted in the APRM cabinet.

Technical

References:

EOP-1A, Step RQA-3 and basis.

R-STM-503, NEUTRON MONITORING INSTRUMENTS SYSTEM, Figure 57 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H4 10CFR Part 55 Content: 55.41(b)(2)

Examination Outline Cross-Reference Level RO 217000 RCIC Tier # 2 Group # 1 Knowledge of the physical connections and/or K/A # K1.04 cause/effect relationships between REACTOR Rating 2.6 CORE ISOLATION COOLING SYSTEM (RCIC) and the following: K1.04 Main condenser Question 60 The RCIC system piping has drain pots located on the RCIC turbine steam supply header that provide protection for the turbine and piping from water hammer. These drain pots are normally ___(1)____ and drain to the _____(2)_____.

A. 1) open

2) suppression pool B. 1) closed

2) suppression pool C. 1) open

2) main condenser D. 1) closed

2) main condenser Answer: C Explanation:

A is wrong due to the second part being incorrect. They are normally open but they drain to the main condenser not the suppression pool.

B is wrong for both parts. They are normally open (not closed) and they drain to the main condenser (not supp pool).

C is correct - they are normally open and drain to the main condenser D is wrong because the first part is incorrect - they are normally open, not closed. The second part is correct.

Technical

References:

STM-0209, revision 12, page 17.

References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0209 N10 Question Source: Bank #

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New X Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO Radiation Control Tier # 3 Group #

Knowledge of radiation exposure limits under K/A # 2.3.4 normal or emergency conditions. Rating 3.2 Question 61 Which of the following exposure limits are Entergy's Routine Annual Administrative Guidelines?

A. TEDE 2000 mrem per year, SDE, WB= 40 rem, and LDE= 12 rem B. TEDE 5000 mrem per year, SDE, WB= 40 rem, LDE= 12 rem C. TEDE 5000 mrem per year, SDE, WB= 50 rem, LDE= 15 rem D. TEDE 2000 mrem per year, SDE, WB= 50 rem, LDE= 15 rem Answer: A Explanation:

A. is correct, the limits are Entergy Routine Annual Administrative Guidelines B. Is incorrect, Maximum Annual Administrative guidelines for Entergy C. is incorrect, they are Annual Regulatory limits D. is incorrect, they are a mix of different limits Technical

References:

EN-RP-201 References to be provided to applicants during exam: None.

Learning Objective: .

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(12)

Examination Outline Cross-Reference Level RO 600000 Plant Fire On Site Tier # 1 Group # 1 Knowledge of the reasons for the following K/A # AK3.04 responses as they apply to plant fire on site: Rating 2.8 Actions contained in the abnormal procedure for plant fire on site Question 62 There is a fire in the D Tunnel RHR B will be placed in alternate shutdown cooling mode Per procedure AOP-0052, Division (a) SRVs are to be used in this situation. The reason they are to be used is (b) .

A. (a) I (b) to ensure that at least eight SRVs are available when control air may be lost for long-term operations B. (a) II (b) because the power supplies to the other division SRVs will be deenergized to remove multiple high impedance faults C. (a) I (b) because the power supplies to the other division SRVs will be deenergized to remove multiple high impedance faults D. (a) II (b) to ensure that at least eight SRVs are available when control air may be lost for long-term operations Answer: D Explanation:

A is wrong. (a) The wrong division of SRVs is provided. See AOP-0052, Section 5.6, sixth bullet. (b) The answer for this part is correct.

B is wrong. (a) The correct division of SRVs is provided. (b) This is incorrect because there are no indications in the question stem that credited power sources for the SRVs have spontaneously deenergized. This distracter deals with the basis for actions taken for procedure step 5.7.1. See the Note before step 5.7.1.

C is wrong because both answers are incorrect. See Explanations for answer A and B.

D is correct. (a) Per AOP-0052, Section 5.6, sixth bullet, in cases where RHR will be used in alternate shutdown cooling mode, it itemizes which SRVs are to be used. The list itemizes each division. For use of RHR B, Division II SRVs are listed. They are available for RPV pressure control, as confirmed on Attachment 1 of AOP-0052. (b) The Note prior to the sixth bullet explains why this action is taken. Fires in various areas can affect the long-term SRV

control air supply (compressed air supply MOV spurious closures). This ensures that there are 8 SRVs available in this condition.

Technical

References:

AOP-0052, Fire Outside the Main Control Room in Areas Containing Safety Related Equipment, Revision 25 OSP-0019, Electrical Bus Outages, Revision 308 System Training Manual R-STM-0109, Main Steam System, Revision 15 SEP-FPP-RBS-002, River Bend Station Fire Fighting Procedure, Revision 2 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b) (5)

Examination Outline Cross-Reference Level RO Emergency Procedures / Plan Tier # 3 Group #

Knowledge of general operating crew K/A # 2.4.12 responsibilities during emergency operations. Rating 4.0 Question 63 In accordance with OSP-0053, Emergency and Transient Response, which of the following EOP procedure steps is considered explicit enough that the skill of the operator alone is all that is required to complete it without hard cards or other procedure guidance?

A. Prevent injection from Low pressure ECCS systems B. Vent the Scram air header C. Initiate injection with ECCS systems which are running but not injecting D. Emergency vent containment Answer: C Explanation:

A is incorrect. This is not listed in the OSP and has an enclosure for it (Encl 27).

B is incorrect. This is not listed in the OSP and has an enclosure for it (Encl 11).

C is correct. This is the only item specifically listed in OSP-0053 that does not require other procedure guidance to complete.

D is incorrect. This is not listed in the OSP and has an enclosure (Encl. 21) for it.

Technical

References:

OSP-0053, revision 23, page 6.

References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)10

Examination Outline Cross-Reference Level RO 205000 Shutdown Cooling Tier # 2 Group # 1 Knowledge of shutdown cooling system (RHR K/A # K4.04 shutdown cooling mode) design feature(s) Rating 2.6 and/or interlocks which provide for the following: Adequate pump NPSH Question 64 The A RHR pump is running in Shutdown cooling mode.

One design feature of the RHR system while in this mode that protects this pump from low Net Positive Suction Head is A. An interlock that trips the pump if valve E-12F010, RHR SDC MAN ISOL VLV, is NOT FULL OPEN.

B. An interlock that trips the pump if Valve SFC-V109, DRYER STORAGE POOL OUTAGE PURIFICATION SYS, is OPEN.

C. An interlock that trips the pump if valve E12-F008A, RHR SHUTDOWN COOLING OUTBD ISOL VALVE, is NOT FULL OPEN.

D. An interlock that trips the pump if Valve E12-F066A, RHR A FUEL POOL COOLING SUCTION, is NOT FULL CLOSED.

Answer: C Explanation:

A is wrong because this valve is a manual valve and does not have an interlock associated with the RHR pump for NPSH/low suction pressure.

B is wrong because this valve has a caution in SOP-0031 to ensure that it is not open to prevent low NPSH but there is no interlock for it with the pump for NPSH/low suction pressure.

C is correct per the Figure 10 of the STM-204 and the SOP-0031 guidance. This valve must be FULL OPEN or the K19A relay energizes and trips the pump due to low NPSH concerns.

D is wrong because the interlock is on the NOT FULL OPEN logic not the NOT FULL CLOSED logic.

Technical

References:

STM-0204, rev 12, page 70, Figure 10, and SOP-0031, revision 327.

References to be provided to applicants during exam: None.

Learning Objective: STM-204 Obj 8 Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)7

Examination Outline Cross-Reference Level RO 295025 High Reactor Pressure Tier # 1 Group # 1 Knowledge of the operational implications of K/A # EK1.01 the following concepts as they apply to high Rating 3.9 reactor pressure: Pressure effects on reactor power Question 65 During an ATWS the following conditions exist:

• Reactor power is stable at 12%

• Reactor pressure is being controlled by the Main Turbine Bypass Valves at

• 920 psig.

• Reactor water level is being maintained at -100 inches How will reactor power be affected if the Main Turbine Bypass Valves failed shut?

Assume no operator actions taken.

A. Reactor power will rise due to a rise in coolant temperature caused by the rise in pressure.

B. Reactor power will be unchanged as the steam line drains control reactor pressure.

C. Reactor power will lower due to the negative reactivity effect of coolant temperature rising.

D. Reactor power will rise due to a reduction in voids caused by rising reactor pressure.

Answer: D Explanation:

A. Incorrect - A rise in coolant temperature would cause a reduction in power.

B. Incorrect - The stem indicates no action has been taken. Operation of the steam line drains requires manual action.

C. Incorrect - The negative reactivity due to voids has a larger affect the moderator temperature.

D. Correct - The closing of the BPVs will cause pressure to reduction in voids in the coolant, resulting in more moderation and higher power.

Technical

References:

R-STM-0509 References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0509 Obj. 15a.

Question Source: Bank # 2010 Question 11

(note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam 2010 -12 Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b) 5

Examination Outline Cross-Reference Level RO 261000 SGTS Tier # 2 Group # 1 Ability to (a) predict the impacts of the K/A # A2.01 following on the STANDBY GAS TREATMENT Rating 2.9 SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low system flow Question 66 Following a LOCA inside containment, the GTS system started automatically. While implementing emergency operating procedures the B train of GTS was placed in standby with Inlet Damper, GTS-AOD1B closed. Later, while LOCA conditions are still present, flow in the A train of GTS lowers to 650 cfm. What is the impact of this condition?

The B train A. will not auto start due to flow not being low enough for the auto start signal.

B. will auto start and reposition the inlet damper automatically.

C. will auto start after the inlet damper has been fully opened manually.

D. can only be started manually after the inlet damper has been fully opened.

Answer: C Explanation:

According to the system training the GTS will only start if the inlet damper is fully open.

There doesnt appear to be any auto opening feature for the inlet dampers. The setpoint for the redundant train to auto start is less than 656 cfm.

A is wrong because the flow is low enough to auto start but the inlet damper is shut B is wrong because there doesnt appear to be an auto open signal for the inlet damper C is correct because see above D is wrong because the GTS can be auto started also.

Other potential distractors:

Will auto start after inlet damper begins to move from its closed seat.

Can only be started manually and the inlet damper will open automatically.

Can only be started manually after the inlet damper begins to move from its closed seat.

Technical

References:

R-STM-257, Standby Gas Treatment System, Rev 5 References to be provided to applicants during exam: None.

Learning Objective: Discuss the operation of the Standby Gas Treatment System including: (3)

a) Automatic initiation/trip signals b) Interlocks/logic of the filter trains/components Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 203000 RHR/LPCI: Injection Mode Tier # 2 Group # 1 Knowledge of electrical power supplies to the K/A # K2.01 following: Pumps Rating 3.5 Question 67 What is the electrical power supply to RHR Pump C?

A. ENS-SWG3A B. ENS-SWG3B C. ENS-SWG1A D. ENS-SWG1B Answer:

Explanation:

A is wrong because the power supply for the C RHR pump is ENS-SWG1B. Plausible if applicant confuses with recirc pump power supplies.

B is wrong because see A C is wrong because the C RHR pump is powered from the ENS-SWG1B. Plausible if applicant believes the C pump is powered from the Division 1 emergency bus.

D is correct because the correct power supply for RHR Pump C is ENS-SWG1B.

Technical

References:

R-STM-0204.012, Residual Heat Removal System (RHR)

R-STM-0300, REV 028, AC Distribution References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(8)

Examination Outline Cross-Reference Level RO 262001 AC Electrical Distribution Tier # 2 Group # 1 Ability to manually operate and/or monitor in K/A # A4.03 the control room: Local operation of breakers Rating 3.2 Question 68 The CRS notifies the Unit Operator that the Control Building operator will be performing a breaker test on ENS-ACB03, E12-C002A RHR A PUMP breaker. The breaker will be in the TEST position with control power fuses INSTALLED. Then the breaker will be CLOSED to support maintenance testing.

Which of the following represents the expected H13-P601 light indications for the RHR A pump breaker when the test conditions mentioned above are established?

A. Red light OFF, Green light OFF, White light OFF B. Red light OFF, Green light OFF, White light ON C. Red light ON, Green light OFF, White light ON D. Red light ON, Green light OFF, White light OFF Answer:

Explanation:

A is wrong because... Red light would be on and white light would be off.

B is wrong because Red light would be on.

C is wrong because White light would be off.

D is correct because Breaker position indication will be available in the MCR when the control power fuses are installed - Red light will be on. The white light however will extinguish if the breaker is not fully racked in to the OPERATE position due to the 52H contact being open.

Technical

References:

ESK-05RHS01 References to be provided to applicants during exam: None.

Learning Objective: HLO-157 Obj 7 & 12; RLP-STM-300 Obj 5 Question Source: Bank # (NRC 2008 Q47) X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam 2008 Q47 Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis

10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 290001 Secondary CTMT Tier # 2 Group # 2 Ability to manually operate and/or monitor in K/A # A4.09 the control room: System status lights and Rating 3.2 alarms: Plant-Specific Question 69 A LOCA is in progress and primary and secondary containment signals are present. The following valves are analyzed in the Main Control Room to ensure correct Auxiliary Building isolation.

1) HVR-AOD164(143), Aux Bldg Inlet Isolation Dampers

2) HVR-AOD18A(B), Aux Bldg to SGT Isolation Dampers

3) HVR-AOD 249, Aux Bldg Outlet Isolation Damper

4) HVR-FN6A(B), Aux Bldg Supply Fans Which one of the following position indication status lights as indicated on the control board is correct for the above valves and fans?

A. 1-Green, 2-Green, 3-Green, 4-Red B. 1-Green, 2-Red, 3-Green, 4-Green C. 1- Green, 2-Green, 3-Green, 4-Green D. 1-Green, 2-Red, 3-Green, 4-Red Answer: B Explanation:

A is wrong because (2) HVR-AOD18A(B), Aux Bldg to SGT Isolation Dampers are open and (4) supply fans are tripped.

B is correct C is wrong because (2) HVR-AOD18A(B), Aux Bldg to SGT Isolation Dampers are open.

D is wrong because (4) supply fans are tripped.

Technical

References:

R-STM-0409, Rev. 6, Auxiliary Building HVAC References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(9)

1. Abnormal Operation (R-STM-0409)

If a Div. 1(2) LOCA (RPV Level 2 or 1.68 psid DW Pressure) signal occurs, is received, the Auxiliary Building Ventilation System automatically isolates the building, and aligns to GTS. When this occurs the following actions take place to provide filtration of the Auxiliary Building atmosphere prior to exhausting it to the Main Plant Exhaust Stack:

• HVR-AOD164(143), Aux Bldg Inlet Isolation Damper, closes,

• HVR-AOD249, HVR-AOD10A(B) & 214(262), Aux Bldg Outlet Isolation Damper, close,

• HVR-AOD18A(B), Aux Bldg to SGT Isolation Damper, opens,

• HVR-FN6A(B), Aux Bldg Supply Fan A(B), trips due to HVR-AOD164(143) damper position,

• HVR-FN7A(B), Aux Bldg Exhaust Fan A(B), trips due to HVR-AOD214(262) or HVR-AOD249(HVR-AOD10A&B) damper position,

• HVR-FN17, D Tunnel Exhaust Fan, trips due to HVR-AOD214(262) damper position and HVR-FN12, D Tunnel Supply Fan, trips due to HVR-FN17 tripping, and

• Standby Gas Treatment Filter Train A(B) starts and draws air from the Auxiliary Building, discharging it to the Main Plant Exhaust Stack.

Examination Outline Cross-Reference Level RO Conduct of Operations Tier # 3 Group #

2.1.39 Knowledge of conservative decision K/A # 2.1.39 making practices. (CFR: 41.10 / 43.5 / 45.12) Rating 3.6 Question 70 In accordance with EN-OP-115, Conduct of operations, Which of the following would be considered a NON-conservative decision making practice?

A. Reducing Reactor Power after a loss of the plant computer B. Validating available information prior to taking any time critical actions C. Re-energizing a vital switchgear during an EOP after a potential fire on the bus D. Consulting an off-shift system engineer before taking action in response to an unexpected system response Answer: C Explanation:

Based on EN-OP-115 Conduct of Operations which describes the thought process utilized for making conservative decisions. Re-energizing a deranged electrical bus, even during the implementation of an EOP would not be prudent without having inspected the bus first.

A. Is wrong because reducing power when indication is unavailable is conservative B. Is wrong because Information should be validated prior to taking action whether it is time critical or not C. Correct D. Is wrong because understanding the system response would ve conservative Technical

References:

EN-OP-115 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H2

10CFR Part 55 Content: 55.41(b)(10)

Examination Outline Cross-Reference Level RO 202002 Recirculation Flow Control Tier # 2 Group # 2 Knowledge of the effect that a loss or K/A # K3.05 malfunction of the RECIRCULATION FLOW Rating 3.2 CONTROL SYSTEM will have on following:

Recirculation pump speed: Plant-Specific Question 71

• The plant is starting up, coming out of an outage where maintenance was conducted on the reactor recirculation system controls

• Per procedure GOP-0001, Attachment 2, Section 4.1, a Control Board Lineup using Attachment 4A of SOP-0003 has been completed

• A control room operator is in the process of starting Reactor Recirculation pump A using SOP-0003, Section 4.4.11

• When the control room operator attempts to depress the RELEASE pushbutton on the STOP/PUSH TO LOCK control switch, the pushbutton will not depress With this malfunction in place, what effect does it have on Reactor Recirculation pump speed control logic?

A. Prevents recirculation pump slow starts only B. Prevents transfers from fast to slow speed C. It would trip the pump if it was running in fast or slow speeds D. It would trip the pump if it was running in fast speed only Answer: D Explanation:

A is wrong. This malfunction would prevent recirculation pump slow and fast starts. See citations in the Explanation for Answer D.

B is wrong. This malfunction would prevent a transfer from slow to fast speed.

C is wrong. This malfunction would not cause a recirculation pump running in slow speed to trip.

D is correct. With the Reactor Recirculation system control board recently lined up according to SOP-0003, Attachment 4A, the PUSH TO LOCK pushbutton on the pumps STOP/PUSH TO LOCK control switch has been depressed. This prevents closing the CB-5 breaker, which limits the speed control functions that can used with the associated recirculation pump

[see Section D.3.b)(1), Page 8 of 76, in R-STM-0053]. Three limitations are in place. Of these, answer D is one of the limitations stated.

Technical

References:

R-STM-0053, Reactor Recirculation System, Revision 14 SOP-0003, Reactor Recirculation System (SYS #053), Revision 313 GOP-0001, Plant Startup, Revision 86

References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 295004 Partial or Total Loss of DC Pwr Tier # 1 Group # 1 Ability to determine and/or interpret the K/A # AA2.01 following as they apply to partial or complete Rating 3.2 loss of D.C. power: Cause of partial or complete loss of D.C. power Question 72 Alarm DIV I 125VDC BUS SPLY/DISTR BRKR OC TRIP is energized.

Which 125VDC circuit breaker trip on Bus ENB-SWG01A could have caused this condition?

A. ENB-ACB561, ENB-BAT01A BATTERY SUPPLY BREAKER B. ENB-ACB583, BACKUP BATTERY CHARGER SUPPLY BREAKER C. ENB-ACB569, ENB-PNL04A PNL04A SUPPLY BREAKER D. ENB-ACB580, BATTERY CHARGER SUPPLY BREAKER Answer: C Explanation:

A is wrong. This DC breaker is associated with ENB-SWG01A, per Attachment 1A of SOP-0049. However, it is not on the list of DC breakers that could cause this alarm in the ARP.

B is wrong. This DC breaker is associated with ENB-SWG01B, and could cause a similar alarm if it opens on overcurrent trip (ARP-808-87, Alarm No. 0940).

C is correct. Per the alarm response procedure (ARP-808-87, Alarm No. 0939), the opening of the DC breaker listed can cause this alarm.

D is wrong. This DC breaker is associated with ENB-SWG01B, and could cause a similar alarm if it opens on overcurrent trip (ARP-808-87, Alarm No. 0940).

Technical

References:

R-STM-0305, DC Distribution, Revision 7 AOP-0014, Loss of 125VDC, Revision 24 ARP-808-87, P808-87 Alarm Response, Revision 24 SOP-0049, 125 VDC System (SYS #305), Revision 35 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2

Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(10)

Examination Outline Cross-Reference Level RO 295007 High Reactor Pressure Tier # 1 Group # 2 Ability to operate and/or monitor the following K/A # AA1.04 as they apply to high reactor pressure: Rating 3.9 AA1.04 Safety/relief valve operation: Plant-Specific Question 73 The reactor is at 100% power when reactor pressure begins to rise to 1140 psig.

Following the initial opening of SRV F051D you would expect to observe F051D cycling between (1) psig and (2) psig.

A. (1) 956 (2) 1063 B. (1) 956 (2) 1103 C. (1) 966 (2) 1063 D. (1) 966 (2) 1103 Answer: A Explanation:

After the initial opening of the SRV the low-low set mode of operation would be active.

F051D is the first SRV to open and, due to low-low set, would cycle between 956 psig and 1063 psig. The 1140 psig in the stem is lower than the setpoint for the second SRV (F051C) to open. The other values (966 and 1103) are the low-low set values for F051C.

Technical

References:

R-STM-0109, Main Steam System, Rev 15 References to be provided to applicants during exam: None.

Learning Objective: List the four modes of SRV operation, including the number SRVs used for each mode. In addition, describe how the SRVs operate in each mode (5).

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)(7)

Examination Outline Cross-Reference Level RO 241000 Reactor/Turbine Pressure Regulator Tier # 2 Group # 2 Ability to (a) predict the impacts of the K/A # A2.02 following on the REACTOR/TURBINE Rating 3.7 PRESSURE REGULATING SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

High reactor pressure Question 74 The plant is operating at 90% power Bypass EHC HCU variable displacement pump D002A is isolated for repairs A reactor transient causes reactor pressure to increase to 1060 psig Main condenser pressure is 23 in. Hg vacuum and slowly lowering Based on the plant configuration, ___(1)__ main turbine bypass valve(s) can open. Alarm response procedures that apply due to the expected alarm(s) direct the operator to

___(2)____.

A. (1) Zero (2) stop all control rod withdrawal B. (1) Two (2) stop all control rod withdrawal C. (1) Zero (2) enter the EOPs D. (1) Two (2) enter the EOPs Answer: B Explanation:

A is wrong. For (1), it is plausible that either the secured bypass EHC HCU pump or vacuum could cause the bypass valves to stay closed. Citing the reference in answer Bs explanation, this is incorrect. (2) is correct.

B is correct. (1) is correct because R-STM-0509, Section F.2.g)(1) (Page 57 of 81) says that for the EHC HPUs, one operating pump will provide adequate EHC fluid pressure for continued operation. Condenser vacuum is lowering, but it is not at the pressure that would preclude bypass valves from opening (8.5 Hg). (2) is correct because the TURBINE BYPASS VALVE OPEN alarm would be lit (ARP-680-07, A07). The Operator Action is to stop all control rod withdrawal.

C is wrong. See Explanations for answers A and D.

D is wrong. (1) is correct, but (2) is incorrect. It is an action for the CONDENSER LO VACUUM alarm (ARP-680-07, D08). Main condenser vacuum isnt low enough to get this alarm, and it isnt low enough to cause a MSIV isolation (8.5 Hg for both).

Technical

References:

System Training Manual R-STM-0509, Electro-Hydraulic Control (EHC), Revision 14 ARP-680-07, P680-07 Alarm Response, Revision 35 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.41(b)(5) 55.43

Examination Outline Cross-Reference Level RO 295038 High Off-site Release Rate Tier # 1 Group # 1 Knowledge of the interrelations between high K/A # EK2.05 off-site release rate and the following: Site Rating 3.7 emergency plan Question 75 The Emergency Plan and emergency procedures contain guidelines for EOP entry. The MINIMUM required EOP-3 entry category for the emergency plan for a high off-site release rate is ____(1)____and the associated instrument used for this determination is ____(2)_____.

A. 1) ALERT

2) RMS-RE125, MAIN PLANT EXHAUST B. 1) ALERT

2) RMS-RE16, CONTAINMENT POST ACCIDENT MONITOR C. 1) SITE AREA EMERGENCY

2) RMS-RE125, MAIN PLANT EXHAUST D. 1) SITE AREA EMERGENCY

2) RMS-RE16, CONTAINMENT POST ACCIDENT MONITOR Answer: A Explanation:

A is correct. EOP-3 entry is for minimum ALERT, and the instrument used is RMS-125.

B is wrong because the second part is wrong (wrong instrument)

C is wrong because the first part is wrong, it is at Alert level, not SAE.

D is correct because both parts are wrong.

Technical

References:

EOP-3, revision 17, EIP-2-001, revision 26.

References to be provided to applicants during exam: None.

Learning Objective: RLP-HLO-0515 Obj 2 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F4 Comprehensive/Analysis 10CFR Part 55 Content: 55.41(b)11

Examination Outline Cross-Reference Level RO Tier # 3 Group #

Ability to use procedures related to shift K/A # 2.1.5 staffing, such as minimum crew complement, Rating 3.9 overtime limitations, etc.

Question 76 Per EN-OP-115, Conduct of Operations, a minimum of (1) members of the site fire brigade is required to fulfill minimum shift staffing requirements. The STA (2) be a member of the fire brigade A. (1) 4 (2) may B. (1) 5 (2) may C. (1) 4 (2) may not D. (1) 5 (2) may not Proposed Answer: D Explanation (Optional):

A. INCORRECT. Part 1 is incorrect but plausible if one believes that the fire brigade can go below minimum staffing for two hours. Part 2 is correct.

B. INCORRECT. According to EN-OP-115, Part 1 is correct as the minimum complement of the fire brigade is five. Part 2 is incorrect but plausible if one believes since the STA is not a licensed position, he can serve on the fire brigade.

C. INCORRECT. Part 1 is incorrect but plausible if one believes that the fire brigade can go below minimum staffing for two hours. Part 2 is incorrect but plausible if one believes since the STA is not a licensed position, he can serve on the fire brigade.

D. CORRECT. According to EN-OP-115, Part 1 is correct as the minimum complement of the fire brigade is five, and part 2 is correct as the STA may not be on the fire brigade.

Technical Reference(s): EN-OP-115, Conduct of Operations, Revision 17 Proposed references to be provided to applicants during examination: None Learning Objective: RLP-HLO-0206, Objective F Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Memory or Fundamental Knowledge F2 10 CFR Part 55 Content: 55.43.5

Comments:

Examination Outline Cross-Reference Level SRO 295004 Partial or Total Loss of DC Pwr Tier # 1 Group # 1 2.2.40 Ability to apply Technical Specifications K/A # 2.2.40 for a system. Rating 4.7 Question 77 During normal plant operations with BYS-TRS4, XFR SW FOR SBO POWER CONNECTION is out of service for repairs, the following alarms are received on H13-P808 Insert 87, including:

• 125VDC BAT CHGR ENB-CHGR1B TROUBLE

• DIV II 125VDC CHGR BRKR ENB-ACB580- OPEN Based on these indications and in order to assess operability of the associated system(s), which of the following conditions should the CRS enter?

A. LCO 3.8.7 Condition A, One inverter on Division I or II inoperable B. LCO 3.8.4 Condition A, One required battery charger on Division I or II inoperable C. LCO 3.8.4 Condition B, Division I or II DC electrical power subsystem inoperable for reasons other than Condition A D. LCO 3.8.9 Condition C, One or more Division I or II DC electrical power distribution subsystems inoperable Answer: B.

Explanation:

A., Incorrect T.S. 3.8.7 Condition A is not correct because B. Correct T.S. 3.8.4 Condition A basis requires the Non safety related Backup charger and SBO Diesel Generator to be available and with BYS-TRS4 out of service the SBO Diesel Generator is NOT available to supply the back up charger, Condition A is applicable.

C. Incorrect because this LCO is for conditions other than Condition A which is not applicable due to the SBO diesel generator not being available.

D. Incorrect LCO 3.8.9 is for DC distribution system during normal operations this distractor is plausible because the candidate may decide the opening of the breaker renders the DC distribution system inoperable which it does not because the system is still capable of supply battery power to the respective loads.

Technical

References:

TS 3.8.4, RLP-STM-0305 References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0305 Obj 7 Question Source: Bank # X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam 2014 -12 NRC Exam (Q76)

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.43(b)2 2

Examination Outline Cross-Reference Level SRO 295024 High Drywell Pressure Tier # 1 Group # 1 Ability to analyze the effect of maintenance K/A # 2.2.36 activities, such as degraded power sources, Rating 4.2 on the status of limiting conditions for operations.

Question 78 3.3.3.1 Post Accident Monitoring (PAM) Instrumentation LCO 3.3.3.1 The PAM instrumentation for each Function in Table 3.3.3.1-1 shall be OPERABLE.

APPLICABILITY: MODES 1 and 2.

ACTIONS

---

NOTE-----------------------------------------------------

Separate Condition entry is allowed for each Function.

COMPLETION CONDITION REQUIRED ACTION TIME A. One or more Functions A.1 Restore required 30 days with one required channel to OPERABLE channel inoperable. status.

B. Required Action and B.1 Initiate action to Immediately associated Completion prepare and submit a Time of Condition A not Special Report.

met.

SR 3.3.3.1.3 (CHANNEL CALIBRATION) was mis-performed, causing PAM Drywell Pressure Instrument ILMS-PT-2A to peg high. 30 days have elapsed and the condition has not yet been corrected.

(1) What Technical Specifications are required to be entered?

(2) When is the Special Report due?

A. (1) TS 3.3.3.1 only (2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> B. (1) TS 3.3.3.1, TS 3.3.1.1 (RPS Instrumentation), TS 3.3.5.1 (ECCS Instrumentation), and others.

(2) 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> C. (1) TS 3.3.3.1 only (2) 14 days D. (1) TS 3.3.3.1, TS 3.3.1.1 (RPS Instrumentation), TS 3.3.5.1 (ECCS Instrumentation), and others.

(2) 14 days Answer: C Explanation:

Part (1) tests on the applicants knowledge of whether the Drywell Pressure PAM instrument uses the same transmitter as RPS and ECCS. RPS and ECCS use 4 transmitters, C71-PT-N050A-D, while PAM uses 2 separate transmitters, ILMS-PT-2A/B. An inoperable PAM drywell pressure transmitter does not affect the RPS/ECCS tech specs.

The tech specs that would be affected by an inoperable RPS-associated drywell pressure transmitter are: TS 3.3.1.1 (RPS Instrumentation), 3.3.5.1 (ECCS Instrumentation), 3.3.6.1 (Primary Containment and Drywell Isolation Instrumentation), 3.3.6.2 (Secondary Containment and Fuel Building Isolation Instrumentation), 3.3.6.3 (Containment Unit Cooler System Instrumentation), 3.3.7.1 (Control Room Fresh Air System Instrumentation). For simplicity, the distractors only name the RPS and ECCS instrumentation tech specs, but acknowledge that others also apply.

Part (2) tests on the applicants knowledge of what is meant by Initiate action to prepare and submit a special report - immediately. If the applicant thinks that this means submit an immediate notification per 10 CFR 50.72, he will choose 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. However, the TS bases explain that this is not a 10 CFR 50.72 report, it is due within 14 days.

From TS 3.3.3.1 Bases:

B.1 If a channel has not been restored to OPERABLE status in 30 days, this Required Action specifies initiation of actions to prepare and submit a Special Report to the NRC. This report discusses the results of the root cause evaluation of the inoperability and identifies proposed restorative actions. The Special Report shall be submitted in accordance with 10 CFR 50.4 within 14 days of entering Condition B. This Action is appropriate in lieu of a shutdown requirement since alternative Actions are identified before loss of functional capability, and given the likelihood of plant conditions that would require information provided by this instrumentation.

A is wrong because the special report is not a 10 CFR 50.72 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report, its due in 14 days.

B is wrong because the special report is not a 10 CFR 50.72 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report, its due in 14 days.

Also, only the PAM TS 3.3.3.1 is affected by the inoperable PT, not the RPS and ECCS tech specs.

C is correct.

D is wrong because only the PAM TS 3.3.3.1 is affected by the inoperable PT, not the RPS and ECCS tech specs.

Technical

References:

R-STM-0057, Rev. 4 (Primary Containment and Auxiliaries)

R-STM-0508, Rev. 6 (RPS)

References to be provided to applicants during exam: None Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 3 10CFR Part 55 Content: 55.43(b)(2)

Examination Outline Cross-Reference Level SRO Radiation Control Tier # 3 Group #

Ability to use radiation monitoring systems, K/A # 2.3.5 such as fixed radiation monitors and alarms, Rating 2.9 portable survey instruments, personnel monitoring equipment, etc.

Question 79 RMS-RE125, MAIN PLANT EXHAUST WRGM has been declared inoperable.

Which of the following is correct regarding meeting the requirements of TRM 3.3.11.3, Radioactive Gaseous Effluent Monitoring Instrumentation?

A. As designed, RMS-RE126, PARTICULATE AND GAS MONITOR satisfies the requirements of TRM 3.3.11.3. No further action is required.

B. RMS-RE126, PARTICULATE AND GAS MONITOR can NOT meet the requirements of TRM 3.3.11.3. Periodic sampling by Chemistry is required.

C. RMS-RE126, PARTICULATE AND GAS MONITOR can satisfy the requirements of TRM 3.3.11.3 if an auxiliary sample holder for particulate and iodine grab samples is placed in line on the particulate and gas skid.

D. RMS-RE126, PARTICULATE AND GAS MONITOR can NOT meet the requirements of TRM 3.3.11.3. Suspend the release of gaseous effluents from this pathway.

Answer: C Explanation:

A. As designed RMS-RE126 does not provide/meet Function 1b and 1c of TRM 3.3.11.3.

B. RMS-RE126 can be used to meet TRM 3.3.11.3. See C.

C. Correct - Per SOP-0086 P&L 2.5, RMS-RE126 can satisfy TRM 3.3.11.3 if a particulate and iodine sampler is utilized on the skid.

D. See C.

Technical

References:

SOP-0086 revision 16, P&L 2.5, TRM 3.3.11.3 rev 5.

References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0511 Obj. 12, 13 Question Source: Bank # 2010S Audit Q97 (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental H4 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)(4)

Examination Outline Cross-Reference Level SRO 295021 Loss of Shutdown Cooling Tier # 1 Group # 1 Ability to determine and/or interpret the K/A # AA2.02 following as they apply to loss of shutdown Rating 3.4 cooling: RHR/shutdown cooling system flow Question 80 The Reactor has been shut down for a refueling outage for 30 days fuel shuffle/reload has been completed.

• The B Train of RHR is in operation removing decay heat.

• Fuel Transfer canal has been drained

• A surveillance is to be performed on the A Train of RHR that will render it inoperable.

• The A train is expected to be inoperable for a total of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Is the removal of one of the required Trains of RHR allowed?

What is the Basis for allowing or not allowing this surveillance to occur?

A. Yes, removal of one required train is allowed per Technical Specifications.

The basis for this allowance is that RCS pressures and Temperatures are being closely monitored as required by LCO 3.4.11, RCS Pressure and Temperature Limits.

B. No, removal of one required train is not allowed per Technical Specifications.

The basis for not allowing this is because decay heat removal is an important safety function that must be accomplished or core damage could result.

C. Yes, removal of one required train is allowed per Technical Specifications.

The basis for this allowance is because the core heat generation can be low enough and the heatup rate slow enough to allow for loss of redundancy in the RHR system.

D. No, removal of one required train is not allowed per Technical Specifications.

The basis for this that two RHR subsystems must remain operable to allow for accurate average coolant temperature monitoring and management of gas voids.

Answer: C Explanation:

A. Is incorrect because this is the basis for securing both trains to preform inservice leak testing and hydrostatic testing.

B. Is incorrect because This is allowed per the note in LCO 3.4.10 C. Correct D. Is incorrect because it is allowed per note in LCO3.4.10 Technical

References:

TS 3.4.10 and Basis

References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental 3 Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)(5)

Note 1 permits both RHR shutdown cooling subsystems and recirculation pumps to be shut down for a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> period. Note 2 allows one RHR shutdown cooling subsystem to be inoperable for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for performance of surveillance tests. These tests may be on the affected RHR System or on some other plant system or component that necessitates placing the RHR System in an inoperable status during the performance. This is permitted because the core heat generation can be low enough and the heatup rate slow enough to allow some changes to the RHR subsystems or other operations requiring RHR flow interruption and loss of redundancy.

Note 3 permits both RHR shutdown cooling subsystems and recirculation pumps to be shut down during performance of inservice leak testing and during hydrostatic testing. This is permitted because RCS pressures and temperatures are being closely monitored as required by LCO 3.4.11.

Examination Outline Cross-Reference Level SRO Radiation Control Tier # 3 Group #

Knowledge of radiation monitoring systems, K/A # 2.3.15 such as fixed radiation monitors and alarms, Rating 3.1 portable survey instruments, personnel monitoring equipment, etc.

Question 81 The surveillance was recently completed for the annulus pressure control system and it has been returned to service. A few hours later, reactor building annulus exhaust radiation monitors RMS-RE11A and RMS-RE11B, both go into HIGH ALARM.

What procedure should the CRS enter first and why?

A. AOP-0003 Automatic Isolations, to verify the Group 8 dampers isolate B. SOP-0059 Reactor Building HVAC, to restore Group 8 dampers from the erroneous isolation C. AOP-0003 Automatic Isolations, to verify the Group 12 dampers isolate D. SOP-0059 Reactor Building HVAC, to restore Group 12 dampers from the erroneous isolation Answer: C Explanation:

A. Incorrect - although AOP-0003 is entered for this condition and must be entered to verify it is valid, the group 8 dampers are for containment purge, not annulus dampers, which are on Group 12 isolations from these two rad monitors.

B. Incorrect - even though this is the procedure used to restore ventilation if the conditions are erroneous, the requirement is to enter the AOP first, then determine if it is erroneous against the values in the Attachments of the AOP, then transition to the SOP to restore the alignment. Also group 8 is not the correct group for these rad monitors.

C. Correct - This is the correct group for these two rad monitors (12), and the AOP is required to be entered to verify it is a valid isolation. SOP is entered if after entering the AOP it is determined it is erroneous, then the CRS would transition to the SOP to restore the alignment of the system dampers.

D. Incorrect - even though this is the procedure used to restore ventilation if the conditions are erroneous, the requirement is to enter the AOP first, then determine if it is erroneous against the values in the Attachments of the AOP, then transition to the SOP to restore the alignment. Group 12 is the correct group for these rad monitors.

E.

Technical

References:

AOP-0003, rev 34. STM 0403

References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0403 Obj 6.3 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.43(b)4

Examination Outline Cross-Reference Level SRO Emergency Procedures / Plan Tier # 3 Group #

Knowledge of operator response to loss of all K/A # 2.4.32 annunciators. Rating 4.0 Question 82 A reactor start-up is in progress and the ATC is raising power with recirc pumps from 50%

power to 60% power when the following annunciators come in:

P680-07A-C07, ANNUNCIATOR LOSS OF DC POWER P630 P680-08A-A04, ANNUNCIATOR LOSS OF DC POWER P850 The CRS will_________________.

A. declare an ALERT within 15 minutes B. declare a NOUE within 15 minutes C. direct the ATC to insert a SCRAM D. direct the UO to start both EDGs Answer: B Explanation:

A. Incorrect - an ALERT would be declared if compensatory instruments were also lost during this event, and there is no other info given that this occurs in the stem.

B. Correct - NOUE is correct based on AOP-0055 and EIP-2-001 for loss of all annunciators without a corresponding loss of comp instruments with no significant transient in progress.

C. Incorrect - this is in direct conflict with AOP-0055 D. Incorrect - this is in direct conflict with AOP-0055 Technical

References:

EIP-2-0001, rev 26, page 108, AOP-0055, rev 21.

References to be provided to applicants during exam: None.

Learning Objective: RLP_OPS-0547, Obj 2, 3 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No

Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.43(b)5

Examination Outline Cross-Reference Level SRO 295005 Main Turbine Generator Trip Tier # 1 Group # 1 Ability to determine and/or interpret the K/A # AA2.05 following as they apply to main turbine Rating 3.9 generator trip: Reactor Power Question 83 According to the basis for TS 3.3.1.1, the Turbine Stop Valve Closure function provides the primary reactor scram signal for a ___(1)___ event, and ensures that the ___(2)___ Safety Limit is not exceeded.

A. (1) Turbine Trip; (2) Reactor Coolant Pressure B. (1) Generator Load Rejection; (2) Reactor Coolant Pressure C. (1) Generator Load Rejection; (2) Minimum Critical Power Ratio D. (1) Turbine Trip; (2) Minimum Critical Power Ratio Answer: D Explanation:

A is wrong because... the Turbine Stop Valve Closure function protects against violating the MCPR safety limit. Plausible due to the potential pressure increase following a turbine trip.

B is wrong because the Turbine Stop Valve Closure function provides the primary scram signal for the turbine trip event. Plausible because the Turbine Control Valve Fast Closure function provides protection against a Generator Load Rejection event. The Turbine Stop Valve Closure function protects against violating the MCPR safety limit. Plausible due to the potential pressure increase following a turbine trip.

C is wrong because the Turbine Stop Valve Closure function provides the primary scram signal for the turbine trip event. Plausible because the Turbine Control Valve Fast Closure function provides protection against a Generator Load Rejection event.

D is correct because per the TS basis, the Turbine Stop Valve Closure function is the primary scram signal for a turbine trip event and protects against exceeding MCPR.

Technical

References:

TS 3.3.1.1 Bases References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0508, Rev. 03, Objective #3 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X

Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)(2)

Examination Outline Cross-Reference Level RO 295016 Control Room Abandonment Tier # 1 Group # 1 Knowledge of the emergency plan K/A # 2.4.29 Rating 4.4 Question 84 The control room is evacuated due to a fire at 1004. The ATC mans the Div I Remote Shutdown Panel. At 1022, the ATC reports that RCIC initiation has just commenced, and that RPV level is being maintained above -160 inches, and RPV pressure is being maintained less than 1094.7 psig.

The SRO should declare a(n):

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Proposed Answer: C Explanation (Optional):

A. INCORRECT.

B. INCORRECT.

C. CORRECT. AOP-0031 requires initiating RCIC for RPV Pressure and Level control within 15 minutes. EIP-2-001 classifies a site area emergency, HS3, for Control room evacuation having been initiated, and control of the plant cannot be established in accordance with AOP-0031, Shutdown from Outside the Main Control Room, within 15 minutes. An Alert is also applicable, HA4, for the fire, but the higher classification of the Site Area Emergency bounds.

D. INCORRECT.

Technical Reference(s):

AOP-0031, Shutdown from Outside the Main Control Room, Revision 323; EIP-2-001, Classification of Emergencies, Revision 026 Proposed references to be provided to applicants during examination: EAL table only.

Learning Objective:

Question Source: New Question History: Last NRC Exam N/A Question Cognitive Level: Comprehension or Analysis 3 10 CFR Part 55 Content: 55.43(b)5

Comments:

Examination Outline Cross-Reference Level SRO Emergency Procedures / Plan Tier # 3 Group #

Ability to take actions called for in the facility K/A # 2.4.38 emergency plan, including supporting or acting Rating 4.4 as emergency coordinator if required.

Question 85 The plant is operating at 100% power when the control room receives the following reports:

• 0800: Field operator makes the report of a fire in the fuel building.

• 0810: Fire Brigade Leader reports that additional outside assistance will be required to fight the fire.

• 0815: Fire Brigade Leader reports that the fire continues to burn, however no safety-related equipment has been affected.

• 0823: The Fire Brigade Leader reports damage to Fuel Pool Cooling Pump, P1A. At the same time, the Shift Communicator reports to the Shift Manager that the initial notification form is ready to be transmitted.

Assuming that it takes the Shift Communicator 5 minutes to prepare or update the notification form, the Shift Manager/Emergency Director should take the following actions:

A. Declare an UE at time 0815. Declare an Alert at 0823. Transmit the UE notification by 0830 and send an Alert notification by time 0838.

B. Declare an UE at time 0815. Declare an Alert at 0823. Direct the Shift Communicator to update the notification form with the Alert declaration and transmit it before 0830.

C. Declare an UE at time 0810. At 0823, declare an Alert and direct the Shift Communicator to cancel the UE notification. Send an Alert notification by 0838.

D. Declare an UE at time 0810. Direct the Shift Communicator to submit the UE notification by 0825. Declare an Alert at 0823 and send an Alert notification by 0838.

Answer: D Explanation:

A is wrong because...the UE should be declared at 0810 when it is known that the fire will not be stopped within 15 minutes. The second half is also incorrect, as there would be sufficient time to update the initial notification to an Alert level prior to sending it at 0830.

B is wrong because the UE should be declared at 0810 when it is known that the fire will not be stopped within 15 minutes. The second half would be correct if 0815 was the correct UE declaration time.

C is wrong because the UE notification should not be cancelled. Since there is not sufficient time to update the UE notification to an Alert, the UE notification should be transmitted and an Alert notification should be transmitted by 0838.

D is correct because the UE declaration time is correct and the notifications are correct.

Technical

References:

EIP-2-001 CLASSIFICATION OF EMERGENCIES EIP-2-002 CLASSIFICATION ACTIONS References to be provided to applicants during exam: EAL Charts Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H4 10CFR Part 55 Content:

55.43(b)(5)

Examination Outline Cross-Reference Level SRO Equipment Control Tier # 3 Group #

Knowledge of conditions and limitations in the K/A # 2.2.38 facility license. Rating 4.5 Question 86 Which of the following Limiting Conditions for Operation (LCOs) listed below provide guidance concerning a supported system LCO not being met because the support system LCO is not met?

A. LCO 3.0.4 B. LCO 3.0.5 C. LCO 3.0.6 D. LCO 3.0.7 Answer:

Explanation:

A. Incorrect - this is for mode changes and LCOs B. Incorrect - this is for equipment returned to service administratively for testing C. Correct - this is for support system LCOs D Incorrect - this is for special operations LCOs Technical

References:

TS Section 3, pages 3.02 and 3.03, Amendment number 173.

References to be provided to applicants during exam: None.

Learning Objective: RLP-HLO-416 Obj 15 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis F3 10CFR Part 55 Content: 55.43(b)1

Examination Outline Cross-Reference Level SRO 264000 EDGs Tier # 2 Group # 1 Knowledge of the operational implications of K/A # 2.4.20 EOP warnings, cautions, and notes. Rating 4.3 Question 87 The plant is operating at 100% power with the Division II Standby Diesel Generator running in parallel with offsite power when a loss of offsite power occurs. Upon the loss of offsite power, the Division II EDG trips on a spurious high jacket water temperature.

The CRS should:

A. Direct implementation of OSP-0053, Attachment 2B, Initiating Division II Standby Diesel Generator Hard Card, to emergency start the Division II diesel generator. Ensure that jacket water temperatures do not exceed a maximum of 186°F to prevent damage to the diesel.

B. Direct implementation of OSP-0053, Attachment 2B, Initiating Division II Standby Diesel Generator Hard Card, to emergency start the Division II diesel generator. Ensure that jacket water temperatures do not exceed a maximum of 200°F to prevent damage to the diesel.

C. Direct operators to place the High Temperature Trip Bypass switch to BYPASS and transition to SOP-0053, Standby Diesel Generator and Auxiliaries, to start the Division II EDG. Ensure that jacket water temperatures do not exceed a maximum of 186°F to prevent damage to the diesel.

D. Direct operators to place the High Temperature Trip Bypass switch to BYPASS and transition to SOP-0053, Standby Diesel Generator and Auxiliaries, to start the Division II EDG. Ensure that jacket water temperatures do not exceed a maximum of 200°F to prevent damage to the diesel.

Answer: D Explanation:

A is wrong because the high temp bypass switch must be taken to bypass. Plausible if applicant believes that the diesel does not trip on high jacket temp when started in emergency mode.

B is wrong because see A.

C is wrong because see D. Plausible if applicant believes that operation above 186F is prohibited.

D is correct because AOP-0004, Loss of Offsite Power, directs that the high temp trip bypass must be taken to bypass and then the diesel restarted IAW SOP-0053. AOP-0004 note further states that jacket water may be allowed to exceed the alarm setpoint of 186F, but temperatures above 200F could cause damage to the diesel. SRO level question due to selection of appropriate procedures.

Technical

References:

AOP-0004, Loss of Offsite Power, Rev 52 SOP-0053, Standby Diesel Generator and Auxiliaries, Rev 332 OSP-0053, Attachment 2B, Initiating Division II Standby Diesel Generator Hard Card, Rev 23 References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0309S, Rev.6, Enabling OBJ E.a Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H4 10CFR Part 55 Content:

55.43(b)(5)

Examination Outline Cross-Reference Level SRO 212000 RPS Tier # 2 Group # 1 2.2.25 Knowledge of the bases in Tech Specs K/A # G2.2.25 for LCOs and Safety limits Rating 2.5 Question 88 One of the ways that the Reactor Protection System provides protection from neutronic/thermal instabilities is that the Average Power Range Monitor Flow Biased Simulated Thermal Power-High Function allowable values are specified in the COLR and are limited by LCO 3.2.4 by restricting the Fraction of Core Boiling Boundary (FCBB).

The normal and setup values are selected by operator manipulation of a Setup button on each flow control trip reference card.

The normal value provides protection when __(1)__ and with the FCBB limit__(2)___.

A. (1) inside the restricted region (2) required to be met B. (1) inside the restricted region (2) not required to be met C. (1) outside the restricted region (2) required to be met D. (1) outside the restricted region (2) not required to be met Answer: D Explanation:

A. Incorrect because normal (ie setup) value is for protection when outside the restricted region not inside it and with FCBB not required to be met per the bases document.

Both aspects are wrong for this distracter.

B. Incorrect because normal (ie setup) value is for protection when outside the restricted region not inside it. Second part of this distracter is correct.

C. Incorrect because normal (ie setup) value is for protection when outside the restricted region (first part is correct) and with FCBB not required to be met, so the second part of this distracter is wrong per the bases document.

D. Correct - because normal (ie setup) value is for protection when outside the restricted region and with FCBB NOT required to be met per the bases document.

Technical

References:

TS Bases pages3.3-8a and 3.3-8b, Revision 4-8.

References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.43(b)2

Examination Outline Cross-Reference Level SRO 290003 Control Room HVAC Tier # 2 Group # 2 2.2.22 Knowledge of limiting conditions for K/A # 2.2.22 operations and safety limits. Rating 4.7 Question 89 During the rotation from Division 1 HVK to Division 2 HVK, Division 2 failed to start due to the failure of the oncoming service water pump. The attempt to rotate back to Division 1 HVK was also unsuccessful because the oncoming chiller did not start. As the CRS, you enter Technical Specification 3.7.3 for two inoperable control room AC subsystems inoperable. It requires you to verify control room area temperature less than or equal to 104°F.

(1) How do you verify this temperature is being met?

(2) What action is required within 30 minutes for this condition?

A. 1) Use lollipop temperature indicator

2) Prop open the doors at the 116 ft. elevation for the battery rooms B. 1) Use hand held pyrometer

2) Prop open the doors at the 116 ft. elevation for the battery rooms C. 1) Use lollipop temperature indicator

2) Prop open the doors at the 98 ft. elevation for the switchgear rooms D. 1) Use hand held pyrometer

2) Prop open the doors at the 98 ft. elevation for the switchgear rooms Answer: B Explanation:

A. Incorrect-Although there are lollipop indicators in some of the other rooms, the main control room does not have these and a handheld pyrometer must be used. The second part is correct per AOP-0060, step 5.1.6. The battery room doors must be propped open within 30 minutes of this event per this step.

B. Correct - this contains the hand held pyrometer, which per Att 1 of AOP-0060, is what is used to measure MCR temperature, and the battery room doors are propped open per step 5.1.6 of the AOP.

C. Incorrect - Although there are lollipop indicators in some of the other rooms, the main control room does not have these and a handheld pyrometer must be used. The second part is also incorrect per AOP-0060, step 5.1.6. The switchgear room doors must be propped open within 2 hrs of this event per this step.5.1.7.

D. Incorrect - The first part is correct, the hand held pyrometer, which per Att 1 of AOP-0060, is what is used to measure MCR temperature The second part is incorrect per AOP-0060, step 5.1.6. The switchgear room doors must be propped open within 2 hrs of this event per this step.5.1.7.

Technical

References:

AOP-0060, rev 10, page 6 and Attachment 1, page 1.

References to be provided to applicants during exam: None.

Learning Objective: RLP-OPS-710 Obj. 1 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content: 55.43(b)2

Examination Outline Cross-Reference Level SRO 295028 High Drywell Temperature Tier # 1 Group # 1 Ability to determine and/or interpret the K/A # A2.01 following as they apply to high drywell Rating 4.1 temperature: Drywell temperature Question 90 The plant is operating at 100% power with all drywell coolers running. Shortly thereafter the following occurs:

• Half scram on Channel B with a Division II NSSSS isolation

• Standby Gas Treatment B and CMS H2 Analyzer B auto start

• Fuel and Control Building Charcoal Ventilation Treatment B starts

• Control Building HVAC B re-aligns to the Charcoal Ventilation Treatment Train A few minutes later the following annunciator is received on H13-601:

• AIR TEMP MON R608 DRYWELL AMBIENT HIGH TEMP What procedure should the CRS use to combat these events?

A. AOP-0010, Loss of One RPS Bus B. EOP-0002 and Enclosure 20 C. SOP-0060, Drywell Cooling D. AOP-0003, Automatic Isolations Answer: A Explanation:

A. Correct - these conditions indicate a loss of RPS bus B which also causes a loss of two of four unit drywell coolers (UC1B and UC1D) and the corresponding high drywell temperature. This procedure has guidance to allow restoring DW coolers without defeating interlocks.

B. Incorrect - EOP-0002 requires entry at 145F and the alarm comes in at approx.

142F. Also, to install enclosure 20 and defeat the interlocks it must be determined that DW temp cannot be maintained below 145F.

C. Incorrect - Although SOP-0060 has directions for starting unit coolers, UC1F cant be started because it has no power so this is not a good choice for DW cooling and does not address loss of the RPS bus either.

D. Incorrect - Resetting isolations per AOP-0003 would be ineffective without power restored to RPS Bus B.

Technical

References:

AOP-0010, rev 21, EOP-0002, rev 16, SOP-0060, rev 10, and AOP-0003, rev 34.

References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0403 Obj. 10; RLP-STM-0057 Obj. 4a Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.43(b)5

Examination Outline Cross-Reference Level SRO 234000 Fuel Handling Equipment Tier # 2 Group # 2 Ability to (a) predict the impacts of the K/A # A2.01 following on the FUEL HANDLING Rating 3.7 EQUIPMENT; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Interlock Failure Question 91 The Refuel Platform is lowering a fuel bundle into the center of the reactor vessel.

The Safety Travel interlock inadvertently actuates.

(1) What is the impact of this condition on the Refuel Platform?

(2) What action is required to complete the fuel move?

A. (1) Only Bridge movement is prevented. Trolley and Main Hoist movement may continue.

(2) Obtain Refuel Floor SRO permission and use the "Travel Override" button.

B. (1) Only Bridge movement is prevented. Trolley and Main Hoist movement may continue.

(2) Obtain Refuel Floor SRO permission and use the "Hoist Override" button.

C. (1) All Bridge, Trolley, and Main Hoist movements are prevented.

(2) Obtain Refuel Floor SRO permission and use the "Hoist Override" button.

D. (1) All Bridge, Trolley, and Main Hoist movements are prevented.

(2) Obtain Refuel Floor SRO permission and use the "Travel Override" button.

Answer: D Explanation:

A. Incorrect because all Bridge, Trolley and Main Hoist movements are prevented.

B. Incorrect because all Bridge, Trolley and Main Hoist movements are prevented and the Hoist Override button is only used to raise the main hoist up past normal up.

C. Incorrect because the Travel Override button must be used D. Correct Technical

References:

FHP-0003,Refuel Platform Operation, (Rev 36)

Section 6.10.3 and 6.11.5.2 References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0055 Obj 11b

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.43(b)7

Examination Outline Cross-Reference Level SRO 295026 Suppression Pool High Water Temp Tier # 1 Group # 1 Ability to determine and/or interpret the K/A # EA2.01 following as they apply to Suppression Pool Rating 4.2 High Water Temperature: Suppression pool water temperature Question 92 The Plant is operating at 90% power and RCIC is being run in CST to CST mode for a PMT.

The Suppression Pool Average Temperature was recorded as follows:

0800 94.0 °F 0805 96.2 °F 0815 100.5 °F 0818 102.3 °F 0820 105.3 °F 0821 106.0 °F The RCIC turbine was tripped at 0818.

Based on the above data, the Unit Supervisor would enter Tech Spec 3.6.2.1______.

A. Condition A for SP Temp exceeding 100°F at time 0815.

B. Condition C for SP Temp exceeding 105°F at time 0820.

C. Condition A for SP Temp exceeding 100°F at time 0818.

D. Condition C for SP Temp exceeding 105°F at time 0820.

Answer: C Explanation:

While testing 105°F is limit, when testing secured 100°F is limit. Therefore condition C never applies for this event.

A. Incorrect - At 0815 testing is still being performed, so Condition A does not apply. Condition C would apply but the higher limit of 105°F is not reached so this doesnt apply.

B. Incorrect - At 0820 testing is secured so condition C does not apply.

C. Correct - At time 0818 as soon as the pump is secured the LCO Condition A TS limit for 100°F with no testing is immediately met and must be entered.

D. Incorrect - At 0820 testing is secured so condition C does not apply.

Because the above the line information is required to be known from memory for RO and SRO applicants, the only concept this question is really testing is knowing that condition C only applies when testing is going on (adds heat to the Supp pool so higher temp is used during this time). Once you throw out condition C, it only leaves condition A and the 100°F limit is met after testing is secured (immediately).

Technical

References:

TS 3.6.2.1, Amendment 81, page 3-6-33.

References to be provided to applicants during exam: None.

Learning Objective: RLP-STM-0656 OBj H12, L7 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H4 10CFR Part 55 Content: 55.43(b)2

Examination Outline Cross-Reference Level SRO 500000 High CTMT Hydrogen Conc. Tier # 1 Group # 2 Knowledge of EOP mitigation strategies. K/A # 2.4.6 Rating 4.7 Question 93 A LOCA has occurred, which for a time uncovered fuel in the core until adequate core cooling could finally be established. The crew is in EOP-0001 and EOP-0002 and the following containment conditions exist:

Pre-LOCA Containment Temperature 90°F POST-LOCA Containment Pressure 2 psig POST-LOCA Containment Temperature 165°F POST-LOCA Containment Hydrogen 3.6%

When the hydrogen igniters were placed in service, several failed to start.

Based on the conditions above, the Control Room Supervisor should direct the crew to use A. Figure 5 HDOL curve in EOP-0001 to determine if containment should be vented B. Enclosure 21, Emergency Containment Venting, to vent containment C. Enclosure 31, H2 Control Systems, to ensure all H2 removal equipment is running D. SAP-2 to ensure all H2 removal equipment is running Answer: D Explanation:

A. Incorrect - Although the curve is on EOP-0002 and SAP-2, it is not on EOP-0001 and would not be used prior to entering the SAP-2.

B. Incorrect - Although this could be used it is not correct at this point to implement this enclosure C. Incorrect - Although this could be used it is not correct at this point to implement this enclosure.

D. Correct - SAP-2 is required to be entered immediately once 3.5% H2 is reached, as given in the stem. Exit from the EOPs and entry to the SAPs are provided at the hydrogen concentration limit of 3.5% to reinforce the SAP entry required in EOP-1/1A and EOP-4/4A Technical

References:

EPSTG*0002, page B-8-33, rev 17, SAP-2 rev 6.

References to be provided to applicants during exam: None.

Learning Objective:

Question Source: Bank #

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New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content: 55.43(b)5

Examination Outline Cross-Reference Level SRO 262002 UPS (AC/DC) Tier # 2 Group # 1 Ability to (a) predict the impacts of the K/A # A2.02 following on the UNINTERRUPTABLE Rating 2.5 POWER SUPPLY (A.C./D.C.) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

A2.02 Over voltage Question 94 Reference Provided During the performance of STP-302-0102, Power Distribution System Operability Check, the Unit Operator has reported a voltage reading of 123.2 volts AC on VBS-PNL01A. A report from the field confirms 123.2 volts AC at the output of ENB-INV01A.

(1) What is the impact of this condition; and (2) What actions should the CRS direct?

A. (1) ENB-INV01A is inoperable and the associated vital bus distribution system is inoperable; (2) Direct performance of SOP-0048 Section 5.5, Transfer from ENB-INV01A to ENB-INV01A1 supplying VBS-PNL01A.

B. (1) Only ENB-INV01A is inoperable.

(2) Direct performance of SOP-0048 Section 5.5, Transfer from ENB-INV01A to ENB-INV01A1 supplying VBS-PNL01A.

C. (1) Only ENB-INV01A is inoperable.

(2) Direct performance of SOP-0048 Section 5.2 Transferring an ENB Inverter from Normal Operation to Maintenance Bypass.

D. (1) ENB-INV01A is inoperable and the associated vital bus distribution system is inoperable.

(2) Direct performance of SOP-0048 Section 5.2, Transferring an ENB Inverter from Normal Operation to Maintenance Bypass.

Answer: A Explanation:

A. Correct - The surveillance contains acceptance criteria for both the inverter and the vital bus. Both are outside of the given acceptance criteria therefore the surveillance requirement is not met so both components must be declared inop (TS 3.8.7 & 3.8.9). Transferring to ENB-INV01A will satisfy TS 3.8.7 and will correct the voltage condition which will also allow exit of TS 3.8.9).

B. The vital bus distribution system is also inop due to voltage being outside the acceptance criteria. Part 2 is correct.

C. Part 1 is incorrect, see B. Operating an inverter in Manual Bypass does not satisfy TS 3.8.7. The correct action is to transfer to the alternate inverter.

D. Part 1 is correct, but operating an inverter in Manual Bypass does not satisfy TS 3.8.7.

The correct action is to transfer to the alternate inverter.

Technical

References:

STP-302-0102, TS 3.8.7, TS 3.8.9 References to be provided to applicants during exam:

TS 3.8.7, TS 3.8.9, STP-302-0102 Learning Objective: RLP-STM-0300 Obj 10 Question Source: Bank # X (note changes; attach parent) Modified Bank #

New Question History: Last NRC Exam 2014-3 Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)5

Examination Outline Cross-Reference Level SRO 239002 SRVs Tier # 2 Group # 1 Knowledge of the bases in Tech Specs for K/A # 2.2.25 LCOs and Safety Limits Rating 4.2 Question 95 According the Technical Specification bases, the safety mode of the Safety/Relief Valves (S/RVs) is designed to protect against which event?

A. MSIVs close with a high flux scram B. Main generator load reject/generator trip C. Loss of main condenser vacuum D. Main turbine stop valve closure with the failure of bypass valves Answer: A Explanation:

A. Correct - Per Technical Specification Bases Section 3.4.4, the most severe pressure transient which the safety function (mode) of the SRV is designed to protect against is closure of all the MSIVs with the reactor scram signal coming from high flux rather than the MSIV valve closure signal.

B. Incorrect - This answer is incorrect because the generator load reject is not the most severe transient. This event assumes the reactor scrams on stop valve closure initiation. This answer is plausible because the event does cause a pressure and power transient.

C. Incorrect - a loss of condenser vacuum is not the most severe pressure transient. The loss of vacuum causes the turbine to trip; stop valve closure will scram the reactor before the pressure/power become too severe. This answer is plausible because an ATWS with bypass valve failure can result in a pressure perturbation when the main turbine trips.

D. Incorrect - This answer is incorrect because this is not the most severe transient which the SRV safety function is designed to preclude. The stop valve closure will cause a reactor scram and the pressure transient is not the most severe. This answer is plausible because a pressure perturbation will occur as a result of the transient.

Technical

References:

TS Bases, 3.4.4, page B-3.4-18, rev 0.

References to be provided to applicants during exam: None.

Learning Objective: RLP-0109, Obj H16, L7 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H2 10CFR Part 55 Content: 55.43(b)2

Examination Outline Cross-Reference Level SRO 215005 APRM / LPRM Tier # 2 Group # 1 Ability to (a) predict the impacts of the K/A # A2.07 following on the AVERAGE POWER RANGE Rating 3.4 MONITOR/LOCAL POWER RANGE MONITOR SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Recirculation flow channels flow mismatch Question 96 The plant is operating at 65% power. A reactor operator reports that recirculation loop flow mismatch is currently 12% of rated core flow.

The CRS should direct core flow to be reduced to less than or equal to a maximum of

___(1)____ of rated core flow per SOP-0003, Reactor Recirculation System.

According to the basis for TS 3.3.1.1, the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function provides ___(2)____.

A. 1) 5%

2) a licensed operating domain which ensures compliance with General Design Criteria 12, Suppression of reactor power oscillations.

B. 1) 5%

2) local neutronic/ thermal hydraulic limits to prevent exceeding the MCPR safety limit.

C. 1) 10%

2) local neutronic/ thermal hydraulic limits to prevent exceeding the MCPR safety limit.

D. 1) 10%

2) a licensed operating domain which ensures compliance with General Design Criteria 12, Suppression of reactor power oscillations.

Answer: D Explanation:

A is wrong because... the maximum flow mismatch below 70% power is 10%. The second half is correct.

B is wrong because the maximum flow mismatch below 70% power is 10%. The APRM flow-biased signal is an averaged flow signal and provides no local neutronic/thermal hydraulic limits. Plausible if applicant believes individual loop flows are used or LPRMs input to the simulated thermal power high function.

C is wrong because See B.

D is correct because 10% is the mismatch limit below 70% power. The basis states that the APRM flow biased function ensures compliance with GDC 12. SRO level due to half of A2 question requiring TS basis knowledge.

Technical

References:

SOP-0003, Reactor Recirculation System, Rev 313 TS 3.3.1.1 Basis References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)(2)

Examination Outline Cross-Reference Level SRO 201002 RMCS Tier # 2 Group # 2 Ability to recognize system parameters that K/A # 2.2.42 are entry-level conditions for Technical Rating 4.6 Specifications.

Question 97 Consider the Following Plant Conditions:

• Reactor Power 12%

• Power escalation is in progress

• One Rod fails to move when attempting to withdraw from 20 to 24 steps Based on the above conditions alone what Technical specification should be entered and what is its basis?

A. TS 3.1.3 Control Rod Operability, because the capability of inserting the control rods provides assurance that the assumptions for scram reactivity in the DBA and transient analyses are not violated.

B. TS 3.1.6 Control Rod Pattern, because the capability of inserting the control rods provides assurance that the assumptions for scram reactivity in the DBA and transient analyses are not violated.

C. TS 3.1.3 Control Rod Operability, because this assures that the control rod patterns are consistent with the assumptions of the CRDA analyses.

D. TS 3.1.6 Control Rod Pattern, because this assures that the control rod patterns are consistent with the assumptions of the CRDA analyses.

Answer: A Explanation:

A. Correct B. is wrong because TS 3.1.6 is not applicable wit thermal power greater than 10%

plausible because this is the correct basis C. is wrong because this is not the reason in the basis for TS 3.1.3 but plausible because this is the basis for TS 3.1.6 D. is correct because TS 3.1.6 is not applicable with the unit greater than 10% thermal power but plausible because this is the correct basis for TS.3.1.6 Technical

References:

TS 3.1.3 and TS 3.1.6 and the Basis for each References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H3 10CFR Part 55 Content:

55.43(b)(2)

Examination Outline Cross-Reference Level SRO 295020 Inadvertent Cont. Isolation Tier # 1 Group # 2 Ability to interpret and execute procedure K/A # 2.1.20 steps. Rating 4.6 Question 98 A plant transient has resulted in numerous alarms and valve actuations in the Main Control Room, including the following:

• All 4 Division 1 RPS solenoid lights on H13-P680 are EXTINGUISHED

• All 4 Division 2 RPS solenoid lights on H13-P680 are ILLUMINATED

• All Division 1 RWCU containment isolation valves are CLOSED

• All Division 2 RWCU containment isolation valves are OPEN

• All Division 1 CCP containment isolation valves are CLOSED

• All Division 2 CCP containment isolation valves are OPEN

• Both the UP STREAM B33-AOVF019, DIV2 and the DN STREAM B33-AOVF020, DIV 1 Reactor Water Sample Line isolation valves closed

1) Which of the following procedures should the CRS direct for these conditions?

2) Why did the sample line isolation valves close?

A. 1) AOP-0003, Automatic Isolations

2) Inadvertent initiation of the CRVICS BOP LOCA isolation signal B. 1) AOP-0003, Automatic Isolations

2) Inadvertent initiation of the Group E isolation signal C. 1) AOP-0010, Loss of One RPS Bus

2) Power to the main steam line high radiation channels A and C was lost D. 1) AOP-0010, Loss of One RPS Bus

2) Power to the isolation logic for both valves was lost Answer: C Explanation:

C is correct-these indications are for loss of RPS Bus A and the required immediate steps to combat this event are to transfer RPS to the alternate bus, which is contained only in AOP-0010, not AOP-0003. The second part of the question deals with what causes both sample valves to close and loss of RPS A bus also causes this failure because it fails the two MSL high rad monitors associated with these two valves (A and C channels).

A is incorrect-wrong procedure and the second part is also wrong because an inadvertent initiation of a group E signal would not cause all of the conditions in the stem but it is plausible if you dont remember what the groups do and remember that a group E signal causes loss of MSL rad monitors and would cause the valves to shut but for a different reason.

B is incorrect because wrong procedure and wrong reason for part 2. The CRVICS BOP LOCA isolation signal would cause the CCP valves to close, but not the other conditions in the stem.

D is incorrect because the second question answer is wrong. The power to the logic is off of different RPS buses so this is not correct.

Technical

References:

AOP-0010, Loss of One RPS Bus, revision 21.

AOP-0003, Automatic Isolations, Revision 34 References to be provided to applicants during exam: None.

Learning Objective: RLP-OPS-AOP0010 Obj 3 Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis H4 10CFR Part 55 Content:

55.43(b)(5)

Examination Outline Cross-Reference Level SRO 295033 High Secondary Containment Area Tier # 1 Radiation Levels Group # 2 K/A # EA2.03 Ability to determine and/or interpret the Rating 4.2 following as they apply to high secondary containment area radiation levels: Cause of high area radiation Question 99 Per the basis for TS 3.6.4.3, Standby Gas Treatment (SGT) System:

1. The SGT system is designed to ___________.

2. The SGT system charcoal filters are designed to ___________.

A. 1) maintain secondary containment at a negative pressure after a design basis LOCA;

2) remove fine particulate matter B. 1) maintain secondary containment at a negative pressure after a fuel handling accident;

2) remove fine particulate matter C. 1) maintain secondary containment at a negative pressure after a fuel handling accident;

2) remove gaseous elemental iodine D. 1) maintain secondary containment at a negative pressure following a design basis LOCA;

2) remove gaseous elemental iodine Explanation:

A is wrong because...first part is correct. The charcoal filter removes iodine, plausible as the first HEPA filter is designed to remove fine particulates to prevent clogging the charcoal filter.

B is wrong because the TS basis specifically states a design basis LOCA as the accident of concern. Plausible as a fuel handling accident could cause fission product release to secondary containment. The charcoal filter removes iodine, plausible as the first HEPA filter is designed to remove particulates to prevent clogging the charcoal filter.

C is wrong because the TS basis specifically states a design basis LOCA as the accident of concern. Plausible as a fuel handling accident could cause fission product release to secondary containment. Second part is correct.

D is correct because Per the TS basis - the design basis for the SGT System is to mitigate the consequences of a loss of coolant accident. For all events analyzed, the SGT System is shown to be automatically initiated to reduce, via filtration and adsorption, the radioactive material released to the environment. The charcoal filter removes iodine, not particulate.

SRO question due to TS basis knowledge.

Technical

References:

TS 3.6.4.3 Basis References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F2 Comprehensive/Analysis 10CFR Part 55 Content:

55.43(b)(2)

Examination Outline Cross-Reference Level SRO Conduct of Operations Tier # 3 Group #

Knowledge of new and spent fuel movement K/A # 2.1.42 procedures. Rating 3.4 Question 100 River Bend is in a refueling Outage and in the process of placing a fuel bundle in a peripheral location. When approaching the location to place the fuel the bridge should be operated in the (1) mode because approaching a peripheral location in the other mode results in (2) which could potentially cause the bundle to contact the wall.

A. (1) XYZ Manual (2) Bundle Swing B. (1) Automatic (2) Bundle Swing C. (1) XYZ Manual (2) Location Overshoot D. (1) Automatic (2) Location Overshoot Answer: A Explanation:

A. Correct B. is wrong because peripheral bundle moves in the core should be performed in XYZ Manual mode of operation. The speed of automatic bridge operation results in bundle swing due to momentum, which could potentially cause the bundle to contact the wall C. is wrong because peripheral bundle moves in the core should be performed in XYZ Manual mode of operation. The speed of automatic bridge operation results in bundle swing due to momentum, which could potentially cause the bundle to contact the wall D. is wrong because peripheral bundle moves in the core should be performed in XYZ Manual mode of operation. The speed of automatic bridge operation results in bundle swing due to momentum, which could potentially cause the bundle to contact the wall Technical

References:

FHP-0003, REFUEL PLATFORM OPERATION. Rev 35 References to be provided to applicants during exam: None.

Learning Objective: Document learning objective if possible.

Question Source: Bank #

(note changes; attach parent) Modified Bank #

New X Question History: Last NRC Exam No Question Cognitive Level: Memory/Fundamental F3 Comprehensive/Analysis 10CFR Part 55 Content: 55.43(b)(6)