ML11196A192

2011 Quad Cities Nuclear Power Station Initial License Examination Administered Written Examination

ML11196A192
Person / Time
Site:	Quad Cities
Issue date:	06/10/2011
From:	NRC/RGN-III
To:	Exelon Generation Co
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ML11167A122	List: ML11196A192 ML11200A096
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Download: ML11196A192 (201)

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EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 1 ID: QDC.ILT.16419 Points: 1.00 The following conditions exist on Unit 1 during a LOCA:

• Reactor pressure is 560 psig

• Indicated Fuel Zone reactor water level is -142 inches

• Drywell pressure is 5 psig

• Drywell temperature is 183°F

• Both reactor recirc pumps are TRIPPED Using the table provided below, what is ACTUAL reactor water level?

FUEL ZONE CORRECTION ACTUAL LEVEL INDICATED LEVEL (INCHES) (INCHES)

> 800 PSIG > 400 PSIG > 100 PSIG 10 -62 -40 -13 0 -70 -48 -22

-10 -77 -57 -32

-20 -85 -65 -41

-40 -101 -82 -60

-59 -115 -98 -78

-80 -132 -116 -97

-100 -147 -133 -116

-120 -163 -151 -135

-142 -180 -169 -155

-166 -199 -190 -178

-184 -213 -205 -194

-191 -218 -211 -201

-200 -225 -219 -209

-260 -272 -270 -265

-300 -303 -304 -303 A. -93 inches B. -110 inches C. -142 inches D. -169 inches Answer: B Page: 1 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Answer Explanation:

Use the PRESSURE column and cross reference to ACTUAL LEVEL.

Since 560 psig is less than 800 psig but greater than 400 psig, use the >400 psig column.

-142 inches is halfway between -133 and -151, so actual level is halfway between -100 and -120. Therefore, actual water level is -110 inches.

Fuel Zone instruments can only be used to determine Reactor water level when the Reactor Recirculation pumps are OFF and their indicated level has been corrected.

Distractor 1: Plausible if candidate uses the >800 psig column.

Distractor 2: Plausible if candidate assumes that water level indication does not have to be corrected when recirc pumps are off.

Distractor 3: Plausible if candidate reverses indicated and actual water level columns.

Reference:

QCAP 0200-10 Rev 41 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.632037)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

SR-0263-K26 (Freq: LIC=B)

EVALUATE given RPV Instrumentation System indications and various plant conditions and DETERMINE a course of action to correct or mitigate the following abnormal condition(s):

a. Rapid RPV depressurization below 450 psig - (determine correction factor)

b. Normal RPV cooldown - (notching)

c. Elevated drywell/reactor building temperatures - (determine if instruments can be used - QGA Detail A)

d. RPV pressure not at instrument calibration pressure - (correct readings using nomograph)

e. Loss of RPV / loop temperature indications - (determine cooldown rate using saturation pressure/temperature graphs)

f. Using heise gauge for RPV water level indication - (determine level using formulas/chart)

g. Determining reactor/cavity level using QCOP 0201-13 Att.A 295001.AA1.07 (CFR 41.7, 45.6)

Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : Nuclear boiler instrumentation system (RO=3.1 / SRO=3.2)

Page: 2 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 2 ID: QDC.ILT.16495 Points: 1.00 Unit 1 is at 100% power with a normal electric plant lineup.

• Bus 14-1 trips on OVERCURRENT.

Which of the following statements is CORRECT regarding the status of MCC 18/19-5?

A. DE-ENERGIZED because the Bus 18 feed breaker must be MANUALLY closed.

B. ENERGIZED from Bus 18 because it is the NORMAL power supply.

C. ENERGIZED from Bus 19 because it is the NORMAL power supply.

D. ENERGIZED from Bus 18 after a 17-second TIME DELAY to allow the Unit 1 EDG to energize Bus 19.

Answer: D Answer Explanation:

This question presents 3 different conditions as a possible reason for an auto-transfer of MCC 18/19-5.

Normally MCC supplied by bus 19 (29), can be supplied by 18 (28).

With an overcurrent trip of Bus 14-1, the Unit 1 EDG will be prevented from loading onto Bus 14-1. Bus 19 will be de-energized.

If power is lost to MCC 18/19-5(28/29-5) by either opening the feed to 18/19-5(28/29-5) from BUS 19(29) or de-energizing BUS 19(29), MCC 18/19-5(28/29-5) will auto-transfer to BUS 18(28) after a 17-second time delay. The 17-second time delay is provided to allow the EDG time to auto-start and re-energize BUS 14-1(24-1) and BUS 19(29). If power is restored from EDG, the MCC power source will not swap (if within 17 sec).

Distractor 1: Plausible if candidate assumes that there is not an auto-transfer to the alternate (similar to other 480 V buses).

Distractor 2: Plausible if candidate assumes that Bus 18 is the normal power supply for MCC 18/19-5.

Distractor 3: Plausible if candidate does not recognize that the overcurrent condition on Bus 14-1 will lockout the bus, preventing the Unit 1 EDG from energizing Bus 14-1 (and subsequently Bus 19).

Reference:

QCOS 6700-01 Rev 12 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Modified from Quad Cities ILT Exam Bank (QDC.ILT.626339)

Question History: N/A Page: 3 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Associated objective(s):

295003.AK3.01 (CFR 41.5, 45.6)

Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER : Manual and auto bus transfer. (RO=3.3 /

SRO=3.5)

SR-6500-K20 (Freq: LIC=B) Given a 4KV / 480 VAC Distribution Systems operating mode and various plant conditions, EVALUATE the following 4KV / 480 VAC Distribution Systems indications/ responses and DETERMINE if the indication/ response is expected and normal.

a. Transformer 11,12,21,22 current

b. 4KV bus watts, voltages, frequency, current and status lights

c. 480 VAC transformers 1A/10/15/16/17/18/19 (2A/20/25/26/27/28/29) current

d. 480 VAC Bus voltages, current and status lights Page: 4 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 3 ID: QDC.ILT.16421 Points: 1.00 Both Units are at rated power when there is a COMPLETE LOSS of the 13.8 KV system.

Which of the following actions are REQUIRED to be completed within 1 HOUR in accordance with station procedures?

A. Transfer Microwave Building power to MCC 28-3.

B. Restore power to a SBO 125 VDC battery charger for each Unit.

C. Energize the Relay House 480V distribution cabinet from MCC 18-2.

D. Start the Security Diesel Generator and energize the Gatehouse 480V MCC.

Answer: B Answer Explanation:

A loss of 13.8 KV will result in a loss of power to the SBO Building and the SBO 125 VDC chargers. If power is not restored to a SBO battery charger on each Unit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, the ability of the SBO DG to start when required will be jeopardized due to a loss of 125 V battery voltage.

Distractor 1: Plausible because the Microwave Building will be on the natural gas generator. Incorrect because there is not a requirement to transfer power to MCC 28-3 within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Distractor 2: Plausible because one of the redundant power supplies is lost. Incorrect because MCC 18-2 is the normal power supply.

Distractor 3: Plausible if candidate assumes that a loss of 13.8 KV will require the SDG to be started.

Reference:

QCOA 6100-17 Rev 10 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 5 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

295004.AK2.01 (CFR 41.7, 45.8)

Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF D.C.

POWER and the following: Battery charger (RO=3.1 / SRO=3.1)

SR-6620-K26 (Freq: LIC=B)

EVALUATE given key Station Blackout Diesel Generator parameter indications and/or responses depicting a system specific abnormality/failure and DETERMINE a course of action to correct or mitigate the following abnormal conditions):

a. Fuel Oil Transfer Pump Failure

b. Lead Air Start System Failure

c. PLC Failure

d. Loss of PMG

e. Loss of 120 VAC

f. Loss of 480 VAC

g. Loss 13.8KV Transformer T42R-6

h. Loss of Battery Charger 6A (7A)

Page: 6 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 4 ID: QDC.ILT.16422 Points: 1.00 Unit 2 is at 100% power when the main turbine TRIPS due to a loss of main turbine lube oil header pressure.

If RPS fails to scram on the turbine trip, which of the following describes reactor power response immediately following the closing of the main turbine control valves?

A. Reactor power will LOWER due to lowering feedwater flow to the reactor.

B. Reactor power will LOWER due to the water level inside the core lowering.

C. Reactor power will RISE due to steam voids collapsing in the core.

D. Reactor power will RISE due to the loss of feedwater heating.

Answer: C Answer Explanation:

As turbine control valves close, reactor pressure rises. This will collapse voids, add positive reactivity and subsequently increase reactor power. That is why there is an anticipatory scram when control valves are < 90% open to prevent a large increase in reactor power.

Distractor 1: Plausible because a turbine trip results in a lowering of feedflow to the reactor.

Distractor 2: Plausible because a collapse in core voids will result in a lowering level within the core.

Distractor 3: Plausible because a trip of the turbine will result in a loss of feedwater heating and reduced feedwater temperature to the reactor. Incorrect because this will not occur immediately after the turbine trip.

Reference:

UFSAR 15.2 Rev 11 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Modified from River Bend ILT Exam Bank Question History: Original version on 2008 River Bend ILT NRC Exam 10 CFR Part 55 Content: 41(b)(1)

SRO Justification: N/A Comments: None Page: 7 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

295005.AK1.01 (CFR 41.8 to 41.10)

Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP : Pressure effects on reactor power (RO=4.0 /

SRO=4.1)

SR-5600-K22 (Freq: LIC=B)

Given a Main Turbine and Auxiliary Systems operating mode and various plant conditions, PREDICT how system/plant parameters will respond to the following Main Turbine and Auxiliary Systems component or controller failures:

a. Gland seal pressure regulator fails high/low

b. Hood spray valve fails closed

c. Loss of lube oil pressure

d. Turning gear failure to engage Page: 8 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 5 ID: QDC.ILT.16423 Points: 1.00 Unit 1 was at rated power when the SUDDEN PRESSURE RELAY initiated for the Unit Auxiliary Transformer (T-11).

• Local operator actions are being performed per QCGP 2-3, Reactor Scram.

What EFFECT will resetting the Main Generator 86 device have on the plant following the T-11 sudden pressure relay initiation?

A. Generator Exciter and Main Field Breakers auto-close.

B. Stator Cooling Water (SCW) pumps auto-start.

C. Hydrogen Seal Oil pumps auto-start.

D. Main Feed Breakers to Busses 11 and 14 auto-close.

Answer: B Answer Explanation:

This question is modeled after SER 3-10, where operators did not fully understand the operational effects of resetting the Main Generator 86 device following a fault within the Unit Auxiliary Transformer.

Resetting the Generator 86 device removes the trip signal from the stator cooling water pumps and they both start on low pump discharge pressure.

Distractor 1: Plausible because the switches for the field breakers are in normal-after-close.

Distractor 2: Plausible because SCW pumps auto-start.

Distractor 3: Plausible because the switches for the main feed breakers to Bus 11 and 14 are in normal-after-control.

Reference:

QCGP 2-3 Rev 72 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: QDC.LN.378618 Question History: N/A 10 CFR Part 55 Content: 41(b)(10)

SRO Justification: N/A Comments: None Page: 9 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

295006.2.4.35 Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (RO=3.8 / SRO=4.0)

SR-6000-K21 (Freq: LIC=B)

Given a Main Generator System operating mode and various plant conditions, PREDICT how key Main Generator System/ plant parameters will respond to manipulation of the following Main Generator System local/remote controls:

a. Exciter field breaker (local)

b. Main generator field breaker (local)

c. Generator lockout (86) / Generator backup lockout (86B) reset

d. Generator stator volt meter selector

e. Generator GCB control switches (901(2)-8 and 912-2)

f. Generator GCB synchronizing switches (901(2)-8 and 912-2)

g. Manual Generator Trip

h. Voltage Reg Mode Control transfer switch

i. Voltage Reg Auto Setpoint Adjust control

j. Voltage Reg Manual Setpoint Adjust control

k. Generator Field Breaker control switch

l. Generator Governor control switch Page: 10 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 6 ID: QDC.ILT.16425 Points: 1.00 (Steam Tables are provided as a reference for this question.)

Unit 2 was at 100% power when a transient occurred.

• At 1600, the reactor is scrammed.

• At 1602, the highest RPV pressure observed is 1135 psig.

• At 1604, the crew enters QOA 0010-05, Plant Operation with the Control Room Inaccessible".

• At 1613, RPV pressure is 785 psig when a NSO arrives at the 2202-5 instrument rack to locally operating ADS Valves for a reactor cooldown.

Which of the following is the LOWEST pressure the RPV can be lowered to by 1700 and still remain in compliance with Technical Specification limitations?

A. 985 psig B. 500 psig C. 430 psig D. 315 psig Answer: B Answer Explanation:

Based on the 100ºF TS 3.4.9 cooldown limit, the RPV coolant temperature could be lowered to 461ºF or an RPV pressure of 467 psig. The lowest selection that would not violate the limit is 500 psig.

The cooldown rate limit must be applied from the highest reactor pressure observed in the 1-hour period (1135 psig).

Distractor 1: 965 psig is plausible if candidate uses the cooldown rate during in-service testing (20ºF/hr).

Distractor 2: 430 psig is plausible if candidate subtracts 100ºF from normal operating temperature at 1005 psig.

Distractor 3: 305 psig is plausible if candidate subtracts 100ºF from the RPV pressure present when the cooldown is initiated.

Reference:

QOA 0010-05 Rev 24, TS LCO 3.4.9 Amendment 228/223, Steam Tables Reference provided during examination: Steam Tables Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.01760)

Page: 11 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Question History: N/A 10 CFR Part 55 Content: 41(b)(14)

SRO Justification: N/A Comments: None Associated objective(s):

295016.AA2.06 (CFR 41.10, 43.5, 45.13)

Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT : Cooldown rate. (RO=3.3 / SRO=3.5)

SR-0201-K28 (Freq: LIC=B)

STATE and EXPLAIN the reasons for the following Reactor Vessel and Internals operating limits and precautions.

a. Head tensioning limitations

b. Heatup/cooldown rate limits

c. RPV vessel flange/head flange temperatures

d. RPV venting Page: 12 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 7 ID: QDC.ILT.16426 Points: 1.00 Unit 1 and 2 were at 100% power when there was a COMPLETE loss of Service Water.

Complete the following two statements:

• Following a loss of Service Water, __(1)__ will be aligned to the CR HVAC 'B' Train for cooling.

• __(2)__ is required to ALIGN the backup cooling water source to the CR HVAC 'B' Train.

A. (1) Diesel Generator Cooling Water (DGCW)

(2) NO operator action B. (1) Diesel Generator Cooling Water (DGCW)

(2) Placing a cooling water supply selector switch to EMERGENCY C. (1) RHR Service Water (RHRSW)

(2) NO operator action D. (1) RHR Service Water (RHRSW)

(2) Placing a cooling water supply selector switch to EMERGENCY Answer: D Answer Explanation:

RHR Service Water is the backup supply available to the safety-related Control Room HVAC train (Train 'B'). To align RHRSW, the operator must place the 1/2-5741-319, AC Unit B Cooling Water Supply Selector switch to EMERGENCY.

Distractor 1: Combination of distractor 2 and 3.

Distractor 2: Plausible because on a loss of Service Water, DGCW is available to cool the HPCI room cooler.

Distractor 3: Plausible if candidate assumes that operator action is not required to align the backup cooling water to the safety-related HVAC train.

Reference:

QCOA 3900-01 Rev 16 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 13 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

295018.AA1.01 (CFR 41.7, 45.6)

Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Backup systems (RO=3.3 / SRO=3.4)

SR-3900-K26 (Freq: LIC=B)

EVALUATE given key Service Water System parameter indications and/or responses depicting a system specific abnormality/failure and DETERMINE a course of action to correct or mitigate the following abnormal condition(s):

a. Loss of instrument air

b. Loss of service water pressure

c. High service water pressure

d. Service water strainer clogging/high dp preventing shifting Page: 14 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 8 ID: QDC.ILT.16427 Points: 1.00 Unit 1 is at rated power with the Instrument Air Compressors (IAC) in the following lineup:

• 1/2B IAC is RUNNING

• 1A IAC is SHUTDOWN with its 912-1 control switch in Normal-After-Trip

• 1/2 IAC is SHUTDOWN with its 912-1 control switch in Pull-To-Lock What is the FIRST response when Instrument Air pressure drops to the actuation setpoint of annunciator 912-1 D-11, UNIT 1/2B INSTRUMENT AIR LOW PRESSURE and what is the reason for the action?

A. A nitrogen backup supply valve will open to keep the MSIVs open.

B. The 1A IAC will start in an attempt to restore Instrument Air pressure.

C. The Service Air to Instrument Air backup (Little Joe) valve will open in an attempt to restore Instrument Air pressure.

D. The 1/2B IAC dryer bypass valve will open to minimize loading on the Instrument Air system.

Answer: C Answer Explanation:

The setpoint for 912-1 D-11 is 85 psig. At 88 psig, the Service Air to Instrument Air backup valve ("Little Joe") to Unit 1 will open, supplying Service Air to the Instrument Air system.

Distractor 1: Plausible if candidate assumes that nitrogen will provide backup control air to the outboard MSIVs (the nitrogen backup supply isolation valve for drywell pneumatics opens at 82 psig drywell pneumatic header pressure). Incorrect because nitrogen is not ported to the outboard MSIVs (unlike the inboard MSIVs).

Distractor 2: Plausible if candidate assumes that the IAC has an auto-start function.

Incorrect because the IAC will not auto-start with its control switch in N-A-T.

Distractor 3: Plausible because the dryer bypass valve will open on a loss of instrument air. Incorrect because it doesn't open until pressure drops to 80 psig.

Reference:

QOA 912-1 D-11 Rev 4, LN-4701 Rev 10 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Dresden ILT Exam Bank Question History: Dresden 2004 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Page: 15 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

295019.AK3.01 (CFR 41.5, 45.6)

Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR : Backup air system supply: Plant-Specific (RO=3.3 / SRO=3.4)

SRN-4701-K22 (Freq: LIC=B NF=B) Given an Instrument Air System operating mode and various plant conditions, PREDICT how system/plant parameters will respond to the following Instrument Air System component or controller failures:

a. Compressor trip

b. Compressor unloading valve failure (open/closed)

c. Dryer switching failure Page: 16 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 9 ID: QDC.ILT.16491 Points: 1.00 Unit 1 is shutdown with the 'B' loop of shutdown cooling in operation.

• A large leak on the 1B Recirc Pump suction piping develops.

Assuming NO operator action is taken, PREDICT how the 1-1001-29B, LPCI LOOP DOWNSTREAM SV will respond if reactor water level lowers from +30 inches to -61 inches.

The 1-1001-29B, LPCI LOOP DOWNSTREAM SV will...

A. CLOSE and NOT automatically OPEN.

B. CLOSE then automatically OPEN.

C. REMAIN OPEN and CAN NOT be remotely CLOSED.

D. REMAIN OPEN but CAN be remotely CLOSED.

Answer: A Answer Explanation:

On a Group II isolation, the following RHR valves will close:

MO 1(2)-1001-20, RHR SYS TO RW UPSTREAM SV.

MO 1(2)-1001-21, RHR SYS TO RW DOWNSTREAM SV.

MO 1(2)-1001-47, SDC SUCT HDR DOWNSTREAM SV.

MO 1(2)-1001-50, SDC SUCT HDR UPSTREAM SV.

MO 1(2)-1001-29A/B, LPCI LOOP DOWNSTREAM SV WHEN BOTH MO 1(2)-1001-47, SDC SUCT HDR DOWNSTREAM SV AND MO 1(2)-1001-50, SDC SUCT HDR UPSTREAM SV are open.

Injection valves will not open with the -59" signal while in SDC mode until the RESET FOR GRP 2 ISOL VLV1-1001-29B pushbutton is depressed.

Distractor 1: Plausible if candidate does not recognize that the RHR 29B valve will close on a Group 2 isolation signal when in SDC mode of operation.

Distractor 2: Plausible if candidate does not recognize that the Group 2 isolation signal must be manually reset before the 29B valve auto-opens on a LOCA signal and confuses the ability to close the 29B valve with the design of the 28B valve (28B valve is interlocked open for 5 minutes with LPCI injection signal and < 325 psig reactor pressure).

Distractor 3: Plausible if candidate does not recognize that the Group 2 isolation signal must be manually reset before the 29B valve auto-opens on a LOCA signal.

Reference:

QCAP 0200-10 Rev 41, QCOA 1000-02 Rev 16 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 Page: 17 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Question Source: Quad Cities ILT Exam Bank (QDC.ILT.627818)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

SR-1000-K12 (Freq: LIC=I)

DESCRIBE how the RHR system responds to an isolation signal.

295021.AK2.03 (CFR 41.7, 45.8)

Knowledge of the interrelations between LOSS OF SHUTDOWN COOLING and the following: RHR/shutdown cooling (RO=3.6 / SRO=3.6)

Page: 18 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 10 ID: QDC.ILT.16492 Points: 1.00 What is the purpose of placing the REFUEL BRIDGE INTERLOCK BYPASS switch on panel 901(2)-28 in NORMAL when the Reactor Mode Switch is in REFUEL?

A. Prevent DROPPING a fuel assembly into the core.

B. Prevent RAISING a fuel assembly above the normal up position limit.

C. Prevent LOADING a fuel assembly into the wrong location within the core.

D. Prevent REACHING criticality by moving control rods and fuel simultaneously.

Answer: D Answer Explanation:

If the Reactor Mode switch is to remain in the REFUEL position, then the Refuel Bridge Interlock Bypass switch in panel 901-28 must be in normal to activate the refuel interlocks.

This switch allows a control rod block or a refuel bridge stop interlock to prevent adding two forms of reactivity at a time (inadvertent criticality).

Distractor 1: Plausible because the grapple is interlocked to fail in the closed position on a loss of motive air.

Distractor 2: Plausible because there is an interlock on the refuel platform hoist that prevents raising a fuel assembly above the normal up position.

Distractor 3: Plausible because loading a fuel assembly into a wrong location into the core is a serious reactivity management violation.

Reference:

QCFHP 0100-01 Rev 31, LF-0803 Rev 15 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Page: 19 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

295023.AK3.02 (CFR 41.5, 45.6)

Knowledge of the reasons for the following responses as they apply to REFUELING ACCIDENTS : Interlocks associated with fuel handling equipment (RO=3.4 / SRO=3.8)

SRL-0803-K16 (Freq: LIC=B NF=B)

STATE the physical location and DESCRIBE the operation of the following Refuel Platform controls:

a. Refueling Interlock Check pushbutton (901(2)-28) (simulate rod withdrawn)

b. Refuel Bridge Intlk Bypass Switch (Normal/Bypass) (901-28)

Page: 20 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 11 ID: QDC.ILT.16429 Points: 1.00 Given the following conditions on Unit 1:

• A LOCA has occurred.

• The Residual Heat Removal (RHR) system is operating in the LPCI Mode.

• Drywell pressure is 5 psig.

• The RHR LOOP A CONTAINMENT CLG PERMISSIVE SWITCH 17 is in ON.

• The CNMT CLG 2/3 LVL AND ECCS INIT BYP SWITCH 18 is in OFF.

Which of the following additional conditions would PREVENT opening the RHR SPRAY SHUTOFF VLV MO 1-1001-37A?

A. Reactor water level lowers to -160 inches.

B. 250 VDC MCC 1A is de-energized.

C. Drywell pressure lowers to 0.8 psig.

D. MO 1-1001-29A, INBD LPCI INJ VLV opens.

Answer: C Answer Explanation:

The RHR Containment Spray and Cooling valves will close when a LPCI initiation signal is present, (ie. 2.5 psig drywell pressure OR -59 in. level and <325 psig OR -59 in. level for 8.5 min.). These valves may be re-opened if the RHR LOOP A/B CONTAINMENT CLG PERMISSIVE SWITCH 17 is placed in the ON position. Two additional conditions will close these valves even if SWITCH 17 is in the ON position. These are:

Condition 1: RPV water level at or below -191 in. (2/3 core height). This can be overridden and the valves opened by placing SWITCH 17 in the ON position AND the CNMT CLG 2/3 LVL AND ECCS INIT BYP SWITCH 18 in the MANUAL OVERRD position.

Condition 2: LPCI initiation signal present and drywell pressure < 1 psig. There is NO override for this since the LPCI initiation signal seals in. The Containment Spray and Cooling valves cannot be re-opened even if both of the above mentioned keylock switches were placed in the ON and MANUAL OVERRD positions respectively.

If a LPCI initiation signal is present and drywell pressure is < 1 psig, then MO 1-1001-37A cannot be opened.

Distractor 1: Plausible if candidate assumes that you cannot spray the containment when reactor water level is below the top of active fuel (-142"). Incorrect because the interlock does not occur until 2/3 core height (-191").

Distractor 2: Plausible because some RHR valves are powered from MCC 1A. Incorrect because MO 1-1001-37A is powered from MCC 18-1B.

Distractor 3: Plausible because MO 1-1001-28A can only be opened if MO 1-1001-29A is fully closed.

Reference:

QCOP 1000-30 Rev 26 Reference provided during examination: None Page: 21 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Cooper ILT Exam Bank Question History: Cooper 2005 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

295024.EK2.13 (CFR 41.7, 45.8)

Knowledge of the interrelations between HIGH DRYWELL PRESSURE and the following:

Suppression pool spray: Plant-Specific (RO=3.8 / SRO=3.8)

SR-1000-K13 (Freq: LIC=B)

DESCRIBE the interlocks associated with the following RHR and RHRSW components, including purpose, setpoints, and when/how they are bypassed, including jumpers.

a. RHR inboard/outboard injection valves (MO 1001-28A/B and 29A/B)

b. RHR heat exchanger bypass (MO 1001-16A/B)

c. RHR containment cooling valves (MO 1001-23,26,34,36,37)

d. RHR pump start (suction valves/power source/FPC)

e. RHRSW pump/valves

f. RHR MO 1001-7A/B/C/D and MO 1001-43A/B/C/D

g. RHR MO 1001-19A/B, MO 1001-43A/B/C/D, and MO 1001-34A/B Page: 22 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 12 ID: QDC.ILT.16511 Points: 1.00 Unit 1 was operating at rated power when the following transient occurred:

• Main Turbine tripped and Bypass valves failed to open.

• RPV pressure peaked at 1380 psig for 5 seconds then lowered to between 1090 - 1100 psig.

• RPV water level lowered to -140 inches for 10 seconds then restored to above 0 inches.

Select the statements describing the (1) Technical Specification Safety Limit concern; and (2) expected plant status.

A. (1) The Reactor Coolant Pressure Safety Limit was exceeded.

(2) All ECCS pumps are injecting.

B. (1) The Reactor Coolant Pressure Safety Limit was exceeded.

(2) HPCI and RCIC ONLY are injecting.

C. (1) The Reactor Vessel Water Level Safety Limit was exceeded.

(2) All Relief and Safety valves opened and are now closed.

D. (1) The Reactor Vessel Water Level Safety Limit was exceeded.

(2) Both Emergency Diesel Generators are running unloaded.

Answer: B Answer Explanation:

RPV Pressure Safety Limit: Reactor steam dome pressure shall be < 1345 psig. At -59 in. RPV water level, HPCI and RCIC will start and inject, however RPV pressure is too high for Low Pressure ECCS pumps to inject.

Distractor 1: Plausible because the Rx. Coolant SL is exceeded and an ECCS pump (HPCI) is injecting but RPV pressure is too high for Low Pressure ECCS pumps to inject.

Distractor 2: Plausible because Relief and Safety valve response is correct, however RPV water level did NOT lower to -142 inches.

Distractor 3: Plausible because the response of the EDGs is correct, but RPV water level did NOT lower to -142 inches.

Reference:

TS SL 2.1.2 Amendment 237/232, UFSAR Section 5.2 Rev 10 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Modified from Oyster Creek Bank Question History: ILT 09-1 NRC RO Exam (Oyster Creek) 10 CFR Part 55 Content: 41(b)(8-10)

SRO Justification: N/A Comments: None Page: 23 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-0201-K32 (Freq: LIC=B)

Given Reactor Vessel and Internals operability status OR key parameter indications, various plant conditions and a copy of Tech Specs, DETERMINE Tech Spec compliance and required actions, if any.

295025.EK1.05 (CFR 41.8 to 41.10)

Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR PRESSURE : Exceeding safety limits (RO=4.4 / SRO=4.7)

Page: 24 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 13 ID: QDC.ILT.16431 Points: 1.00 Unit 1 was at rated power when a transient occurred.

• Relief valves are automatically CYCLING to control RPV pressure.

The following indications are available for Torus water temperature:

(Located on 901-21 panel) (Located on 901-4 panel)

Given the above plant conditions:

(1) Per QCAP 0200-10, EOP (QGA) Execution Standards, which is the PREFERRED temperature recorder to use when determining if Torus temperature has reached the EOP entry condition; and (2) What operator action can be taken to obtain the BEST possible indication of bulk torus water temperature?

A. (1) TR 1-1640-8 (2) Place RHR in the Torus Cooling mode of operation B. (1) TR 1-1640-8 (2) Use ONLY the 3B and 3C relief valves to maintain RPV pressure C. (1) TR 1-1640-9 (2) Place RHR in the Torus Cooling mode of operation D. (1) TR 1-1640-9 (2) Use ONLY the 3B and 3C relief valves to maintain RPV pressure Answer: C Page: 25 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Answer Explanation:

Response to Torus water temperature should be based upon the bulk temperature of the pool rather than the temperature of an individual bay. To accomplish this, use the following sensors which are listed in order of preference:

(1) TR 1(2)-1640-9, TORUS H2O TEMP CH 1 or 2 (2) Process computer point C169(C269) or C170(C270)

(3) TR 1(2)-1640-200A/B, TORUS H2O TEMPERATURE point 9 (4) Average of CH 1 and 2 on TR 1(2)-1640-8, TORUS H2O TEMP Torus cooling should be run to obtain mixing for the best possible indication of bulk temperature anytime steam is released to the Torus. This ensures no localized hot spots in the Torus water exist.

Distractor 1: Plausible because TR 1-1640-8 has a torus temperature that is higher than TR 1-1640-9 and exceeds the EOP entry condition.

Distractor 2: Combination of distractor 1 and 3.

Distractor 3: Station procedures direct the use of the 3B and 3C relief valves when controlling reactor pressure because they have the 14-second time delay for opening following closure. It is plausible that the candidate assumes that the 3B and 3C relief valves are used to distribute heat evenly in the torus.

Reference:

QCAP 0200-10 Rev 41, QCOP 0203-01 Rev 13 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: Question must be printed in COLOR.

Associated objective(s):

SR-0001-K07 (Freq: LIC=I)

For the following plant parameters, STATE the 'preferred' indication:

a. Torus water temperature

b. Drywell temperature

c. Reactor building temperature 295026.2.2.44 Ability to interpret control room indications to verify the status and operation of system, and understand how operator actions and directives affect plant and system conditions. (RO=4.2 / SRO=4.4)

Page: 26 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 14 ID: QDC.ILT.16499 Points: 1.00 Unit 1 is at full power when a design basis accident (DBA) LOCA occurs in the drywell.

• The Primary Containment responds as designed.

• NO operator action is taken.

Which of the following statements is CORRECT regarding the Torus AIRSPACE temperature as compared to Drywell AIRSPACE temperature shortly after the DBA-LOCA occurs?

A. Torus airspace temperature is CLOSE TO drywell temperature because VACUUM BREAKER operation equalizes airspace temperatures.

B. Torus airspace temperature is CLOSE TO drywell temperature because the DOWNCOMER flow to the Torus equalizes airspace temperatures.

C. Torus airspace temperature is MUCH BELOW drywell airspace temperature because the SUPPRESSION POOL cools the non-condensible gases transferred to the Torus airspace.

D. Torus airspace temperature is MUCH BELOW drywell airspace temperature because TORUS SPRAYS are running to reduce Torus airspace temperature.

Answer: C Answer Explanation:

Immediately following a DBA LOCA, drywell pressure rises and the drywell atmosphere (non-condensible gases and steam) is sent to the suppression pool via the downcomers.

The suppression pool condenses the steam and cools the non-condensible gases.

Therefore, the Torus airspace temperature is much below the drywell airspace temperature (which is primarily a saturated system and at a very high temperature).

Distractor 1: Plausible because the Torus-to-Drywell vacuum breakers will open as drywell atmosphere cools. However, this will not result in equalizing temperatures because the Torus airspace is pulled into the drywell.

Distractor 2: Plausible if candidate does not apply the design and function of the suppression pool.

Distractor 3: Plausible because Torus sprays will be placed in service for this accident.

Incorrect because they will not automatically operate and no operator action has been taken to manually start Torus sprays.

Reference:

UFSAR Chapter 6.2 Rev 11 (Figures 6.2-1, 25a)

Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(10)

Page: 27 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

SRO Justification: N/A Comments: None Associated objective(s):

295028.EA2.06 (CFR 41.10, 43.5, 45.13)

Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE : Torus/suppression chamber air space temperature: Plant-Specific.

(RO=3.4 / SRO=3.7)

SR-TAA-K10 (Freq: LIC=Q) DESCRIBE the initial conditions assumed and the UFSAR predicted plant reponse for the following transients in accordance with the UFSAR:

a. Large recirculation line break

b. Small recirculation line break

c. Steam line break Page: 28 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 15 ID: QDC.ILT.16432 Points: 1.00 Unit 2 has developed an unisolable leak on the ECCS suction header. Suppression pool level continues to LOWER.

• Adequate core cooling is being maintained by the HPCI system.

Which one of the following corresponds to the MINIMUM suppression pool level, if any, at which the HPCI system must be TRIPPED even if core cooling will be lost?

A. 14 feet B. 11 feet C. 5 feet D. HPCI is NOT required to be tripped on low suppression pool level.

Answer: B Answer Explanation:

The HPCI system must be secured when suppression pool level reaches 11 ft. This is because operation below 11 ft will result in the HPCI exhaust line being uncovered and subsequent pressurization of the Primary Containment.

Distractor 1: Plausible because this is the approximate torus level specified for EOP entry.

Distractor 2: Plausible because this is the torus level that prevents relief valve operation due to uncovering the spargers.

Distractor 3: Plausible if candidate assumes that the exhaust from HPCI operation is within the capacity of the primary containment vent (RCIC can be operated regardless of torus water level).

Reference:

QGA 200 Rev 9 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Brunswick ILT Exam Bank Question History: Brunswick 2003 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: Delete QGA 200 statement regarding HPCI operation at 11 ft from any reference materials provided to candidate.

Page: 29 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

295030.EA1.05 (CFR 41.7, 45.6)

Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: HPCI (RO=3.5 / SRO=3.5)

SR-0001-K23 (Freq: LIC=B)

Given QGA 200, 'Primary Containment Control' and QGA 200-5, 'Hydrogen Control',

EXPLAIN the reasons for the actions.

Page: 30 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 16 ID: QDC.ILT.16500 Points: 1.00 Complete the following statement regarding the Automatic Depressurization System (ADS):

The purpose of the ADS logic is to actuate relief valves to...

A. quickly cooldown the reactor pressure vessel within Technical Specification thermal limits.

B. control reactor pressure during transient conditions to prevent the need for safety valve actuation.

C. allow use of low pressure ECCS by reducing reactor pressure during small area breaks.

D. reduce reactor pressure during a LOCA to ensure peak drywell pressure remains below the design limit.

Answer: C Answer Explanation:

Per the UFSAR, the ADS system allows use of LPCI or core spray as a backup to HPCI by depressurizing the reactor pressure vessel for small area breaks.

This is the reason for ADS actuation on low reactor water level. To avoid creating a specific determiner (small break LOCA), the initial condition of low reactor water level condition was omitted from the stem. Instead, the stem is worded in a "system purpose" format.

Distractor 1: Plausible because the ADS will quickly cooldown the reactor. Incorrect because the Technical Specification cooldown rate will be exceeded if ADS actuates.

Distractor 2: Plausible because this is the purpose of the relief valves.

Distractor 3: Plausible because any reduction in RPV pressure using relief valves will reduce peak drywell pressure.

Reference:

UFSAR 7.3.1.4 Rev 10 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.699796)

Question History: N/A 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Page: 31 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-0203-K01b (Freq: LIC=B)

STATE the purpose(s) of the ADS logic system including applicable design bases.

295031.EK3.01 (CFR 41.5, 45.6)

Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL : Automatic depressurization system actuation. (RO=3.9 / SRO=4.2)

Page: 32 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 17 ID: QDC.ILT.16434 Points: 1.00 Unit 1 was at 58% power when a manual scram was inserted due to an equipment malfunction.

The following conditions exist:

• Multiple control rods FAILED to fully insert into the reactor core.

• Reactor power is 7% and steady

• Control rod J-6 is selected and is currently at position 24.

• Control rod J-6 is NOT in the current sequence step.

• The RWM mode switch is in NORMAL.

Complete the following statements.

1.) If the ROD OUT NOTCH OVERRIDE switch is held in the EMERG ROD IN position, control rod J-6 will ____.

2.) When using the Emergency Rod In function of RMCS, the Automatic Sequence Timer ____.

A. (1) insert (2) is bypassed B. (1) remain at position 24 (2) is bypassed C. (1) insert (2) remains enabled D. (1) remain at position 24 (2) remains enabled Answer: B Answer Explanation:

Emergency Rod In should be for Control Rod insertion during emergency conditions, such as an ATWS condition or when use of cram arrays is desired.

"Emergency Rod In" function bypasses all interlocks to insert the rod except the rod worth minimizer insert block and any select blocks. It directly energizes the directional control valves by bypassing the timer. No settle function is provided in this position.

Water is forced past the seals in the CRD while settling into a notch.

Distractor 1: Plausible if candidate assumes that the RWM insert block is bypassed using the emergency rod in position.

Distractor 2: Combination of distractor 1 and 3.

Distractor 3: Plausible if candidate assumes that the timer is NOT bypassed with emergency rod in (similar to the functionality of "continuous rod in" position).

Reference:

QCOP 0280-01 Rev 16 Reference provided during examination: None Cognitive level: High Page: 33 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.626954)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

SR-0280-K21 (Freq: LIC=B)

Given a Reactor Manual Control System (RMCS)/ Rod Position Information System (RPIS) operating mode and various plant conditions, PREDICT how key system/ plant parameters will respond to manipulation of the following Reactor Manual Control System (RMCS)/ Rod Position Information System (RPIS) controls:

a. Rod Select Matrix (1) Rod Select Matrix power switch (2) Rod Select switches

b. Rod Movement switches (1) Rod Movement Control Notch In/Out switch (2) Rod Out Notch Override switch (3) Emergency In (4) Rod Drift Alarm Test switch (5) Timer Malfunction Select Block Test/Reset switch 295037.EK2.11 (CFR 41.7, 45.8)

Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following:

RMCS: Plant-Specific (RO=3.8 / SRO=3.9)

Page: 34 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 18 ID: QDC.ILT.16436 Points: 1.00 What is the basis for maximizing Turbine Building ventilation when executing QGA 400, Radioactivity Release Control?

A. To provide dilution flow for elevated releases from the SBGTS through the Main Chimney.

B. To allow operation of Turbine Building equipment without exceeding max safe temperature conditions.

C. To allow personnel access to the Turbine Building and discharge radioactivity through an elevated, monitored release point.

D. To maintain the Secondary Containment differential pressure within operational limits.

Answer: C Answer Explanation:

Turbine Building ventilation is maximized to allow personnel to access the Turbine Building. This is essential for responding to emergencies or transients which may degrade into emergencies. These buildings are not always airtight structures, and radioactivity released inside the buildings would not only limit personnel access, but would eventually lead to an unmonitored ground level release.

A monitored, elevated release (vice unmonitored ground release) is necessary to ensure the protection of the general public.

Distractor 1: Plausible because Turbine Building ventilation will provide some dilution flow, however, the concern for QGA 400 is release outside secondary containment; diluting SBGTS flow would not address the problem QGA 400 is attempting to address.

Distractor 2: Plausible because there are EOPs that address the maximum temperature in the Reactor Building. However, Turbine Building temperatures are not addressed in QGA 400.

Distractor 3: Plausible because this is the basis for restarting Reactor Building Ventilation if the restart will not result in an excessive release of radioactivity to the environment. Incorrect because maximizing Turbine Building ventilation will not assist with Secondary Containment.

Reference:

L-QGA400 Rev 7 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.15618)

Question History: N/A 10 CFR Part 55 Content: 41(8-10)

SRO Justification: N/A Page: 35 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Comments: None Associated objective(s):

SR-0001-K35 (Freq: LIC=B)

Given QGA 400, 'Radioactivity Release Control', EXPLAIN the reasons for the actions.

295038.EK1.02 (CFR 41.8 to 41.10)

Knowledge of the operational implications of the following concepts as they apply to HIGH OFF-SITE RELEASE RATE : Protection of the general public (RO=4.2 /

SRO=4.4)

Page: 36 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 19 ID: QDC.ILT.16437 Points: 1.00 Unit 1 is operating at 100% power when a FIRE occurs in the Bus 11 and 12 area.

• A manual reactor SCRAM is inserted and reactor water level drops to -35 inches.

Which one of the following describes the correct Operator response?

A. Complete QCARP immediate operator actions first until the plant is in a safe condition, then implement QGA 100 (RPV Control).

B. Implement QGA 100 (RPV Control) and QGA 300 (Secondary Containment Control) concurrently.

C. Complete QGA 100 (RPV Control) actions first until RPV water level is restored, then implement QCOA 0010-12 (Fire/Explosion).

D. Implement QGA 100 (RPV Control) and QCOA 0010-12 (Fire/Explosion) concurrently.

Answer: D Answer Explanation:

Per QCAP 0200-10, Emergency Operating Procedure (QGA) Execution Standards:

"Concurrent execution of the QGA procedures is necessary to control plant conditions.

The operating crew is expected to monitor and maintain all key plant parameters as directed by the QGA procedures. In order to successfully accomplish this, multiple procedures will need to be executed concurrently. Included within the QGA procedures are references to additional plant operating procedures. These references are provided as an aid to the operator indicating additional procedures which may be referenced to perform the task. Procedures referenced by the QGAs are not all inclusive of the plant procedures which may be employed as a plant transient condition is handled or direction that a specific procedure must be used."

Both the EOP and QCOA are required to be executed concurrently to maintain plant parameters.

Distractor 1: Plausible if candidate has a misconception on the use of QCARP (Appendix R) fire procedures. Incorrect because QCARP actions are taken if EOP actions are insufficient or inadequate.

Distractor 2: Plausible if candidate assumes that the fire is in secondary containment and requires QGA 300 actions.

Distractor 3: Plausible if candidate assumes that QGA actions must be completed prior to taking QCOA actions.

Reference:

QCAP 0200-10 Rev 41 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 Page: 37 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Question Source: DC Cook ILT Exam Bank Question History: DC Cook 2008 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(10)

SRO Justification: N/A Comments: This is a RO level question because it can be answered by knowing the purpose and overall mitigative strategy of each procedure. The facility objective also supports screening this question as a RO level question.

Associated objective(s):

SR-0001-K05 (Freq: LIC=I)

DESCRIBE the procedure use guidelines contained in QCAP 0200-10

a. QGA procedure entry

b. QGA execution / concurrent use (QGA, QCOA, QCOP, and QCGP)

c. QGA procedure exit 600000.2.4.08 Knowledge of how abnormal operating procedures are used in conjunction with EOPs. (RO=3.8 / SRO=4.5)

Page: 38 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 20 ID: QDC.ILT.16438 Points: 1.00 Unit 1 was at 100% power when grid conditions resulted in LOW switchyard voltage.

The Unit 1 Main Generator is currently operating with the following parameters:

• 60 psig hydrogen pressure

• Generator voltage regulator in Auto Which of the following operating conditions, if any, are OUTSIDE of the Main Generator Capability Curve limitations?

Condition 1: 17.1 kV at 930 MWe and 250 MVAR Condition 2: 17.1 kV at 930 MWe and 100 MVAR Condition 3: 18.0 kV at 910 MWe and 100 MVAR GENERATOR CAPABILITY CURVES AND MINIMUM EXCITATION LIMITER SETTING Regulator: In Service 500.00 450.00 400.00 Gross Reactive Output in MVar (Lagging)

Et = 17.1 kV [105%]

350.00 Et = 18.0 kV [100%]

300.00 Et = 18.9 kV [95%]

250.00 Capability Curve @ 60 psig H2 200.00 0.945 Power Factor (PF) Line (Unit 1 only: Maintain 150.00 Main Generator FIELD CUR less than or equal to 3679 Amps if possible OR stay below line.)

100.00 50.00 0.00 900 910 920 930 940 950 960 970 980 990 1000 1010 1020 1030 1040 1050 1060 Gross Power Output in MWe A. Condition 1 only.

B. Condition 2 only.

C. Condition 1 and 3 only.

D. ALL conditions are within operating limits.

Answer: B Page: 39 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Answer Explanation:

Generator operation must be below the capability curve at 60 psig and above the Minimum Excitation Line (MEL) for terminal voltage on Limiter #1.

Using Attachment B Page 2, plot the position of the generator operation at 930 MWe and 100 MVARs. Operation of the generator must be above the terminal voltage line, therefore 18.0 kV is the minimum terminal voltage value listed that meets this criteria. A terminal voltage of 17.1 kV is too low.

Distractor 1: Plausible if candidate assumes that you must be above the most restrictive MEL line (17.1 kV).

Distractor 2: Plausible if candidate assumes that you must be below the MEL.

Distractor 3: Plausible if candidate assumes that you only have to be below the hydrogen pressure line of 60 psig.

Reference:

QCGP 3-1 Rev 63 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Associated objective(s):

SR-6000-K34 (Freq: LIC=I)

Given main generator MWe, MVARS, and hydrogen pressure DETERMINE if main generator is within the limits of the Main Generator Reactive Capability Curve.

700000.AA2.02 (CFR: 41.5 and 43.5 / 45.5, 45.7, and 45.8)

Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Voltage outside the generator capability curve (RO=3.5 / SRO=3.6)

Page: 40 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 21 ID: QDC.ILT.16439 Points: 1.00 Unit 1 was at 100% reactor power when a leak developed in the Main Condenser boot.

• A manual Reactor scram was inserted and the Reactor Mode switch was placed in SHUTDOWN.

• Reactor pressure is 890 psig and slowly rising.

• Main Condenser backpressure is currently 7" Hg and rising at a rate of 2" Hg per minute.

If backpressure continues to rise at this rate, which of the following describes RPV pressure control three (3) minutes later?

(Assume NO further operator action is taken.)

A. Turbine Control Valves and Bypass Valves will be closed and RPV pressure will rise until relief valves open.

B. Turbine Control Valves will open to control RPV pressure.

C. Bypass Valves will open to control RPV pressure.

D. Turbine Control Valves and Bypass Valves will be closed and RPV pressure will rise until safety valves open.

Answer: C Answer Explanation:

7" Hg + ((3 minutes) X (2" Hg/minute)) = 13" Hg backpressure The turbine trips at 10" Hg backpressure and the Bypass valves will be interlocked closed at 23" Hg backpressure. Therefore, the bypass valves will be available to control RPV pressure. Decay heat will cause Reactor pressure to rise until the pressure reaches pressure set (920 psig).

Distractor 1: Plausible if candidate assumes that the bypass valves are not available for pressure control.

Distractor 2: Plausible if candidate does not recognize that the turbine trips and control valves close at 10" Hg backpressure.

Distractor 3: Plausible if candidate assumes that the main condenser is not available as a heat sink and decay heat production will exceed the capacity of the relief valves.

Reference:

QOA 3300-02 Rev 37 Reference provided during examination: N/A Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 2 Question Source: Modified from Quad ILT Exam Bank (QDC.ILT.15477)

Question History: Original version of question on Quad 2009 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(10)

Page: 41 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

SRO Justification: N/A Comments: None.

Associated objective(s):

SR-3200-K24 (Freq: LIC=B)

Given a Condensate/Feedwater System operating mode and various plant conditions, PREDICT how each supported system will be impacted by the following Condensate/Feedwater System failures:

a. Loss of condensate reject flow

b. Feed header rupture inside the drywell

c. Loss of vacuum

d. Condenser tube rupture 295002.AA2.01 (CFR 41.10, 43.5, 45.13)

Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM : Condenser vacuum/absolute pressure (RO=2.9 / SRO=3.1)

Page: 42 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 22 ID: QDC.ILT.16440 Points: 1.00 Unit 1 was at rated power when a loss of feedwater occurred.

• Reactor is SCRAMMED.

• HPCI is manually started and is injecting into the reactor.

• Reactor water level is -23 inches and rising.

If NO operator action other than starting HPCI was taken, which of the following RWCU indications is EXPECTED?

A. Red ON light is LIT for PRECOAT PUMP 1-1279-7 B. Red OPEN light is LIT for the DEMIN BYP VLV MO 1-1201-133 C. Amber TRIP light is LIT for RWCU Recirc Pumps 1-1205A and 1-1205B D. Flowrate of 230 gpm is indicated on FI 1-1290-30A(B), A(B) DEMIN FLOW Answer: C Answer Explanation:

A low reactor water level condition results in a PCIS Group 3 isolation. As soon as the MO 1-1200-2 and 5 valves (inboard and outboard suction isolation valves) are not full open, the RWCU recirc pumps automatically trip.

Distractor 1: Plausible if candidate assumes that the precoat pump auto-starts to maintain precoat on the RWCU filter demins. Incorrect because that is the function of the hold pump.

Distractor 2: Plausible if candidate assumes that the demin bypass valve opens to establish a flow path for the RWCU recirc pumps. Incorrect because the hold pump recycle valve opens.

Distractor 3: Plausible if candidate assumes that the hold pump will maintain demin flow at normal flowrates. Incorrect because the hold pump auto-starts around 60 gpm demin flow.

Reference:

QCOP 1200-11 Rev 33 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 2 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 43 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-1200-K12 (Freq: LIC=I)

DESCRIBE how the Reactor Water Cleanup System responds to the following isolations.

a. System isolation

b. Filter-demin isolation

c. Reject line isolation 295009.AA1.04 (CFR 41.7, 45.6)

Ability to operate and/or monitor the following as they apply to LOW REACTOR WATER LEVEL : Reactor water cleanup (RO=2.7 / SRO=2.7)

Page: 44 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 23 ID: QDC.ILT.16441 Points: 1.00 Which of the following describes the REASON for monitoring Drywell radiation levels during accident situations with elevated Drywell pressure?

Drywell Radiation Monitor readings are used by Control Room Operators to determine...

A. offsite release rate.

B. if the core is submerged.

C. if core damage has occurred.

D. if the Reactor is shutdown following a scram.

Answer: C Answer Explanation:

Per QCAP 0200-10, EOP (QGA) Execution Standards, core damage is evaluated by monitoring primary containment hydrogen and radiation values and trends.

Distractor 1: Plausible because the determination of noble gas offsite release rate utilizes radiation levels.

Distractor 2: Plausible because if fuel damage may occur is the core is not submerged.

Incorrect because core submergence is based on water level and ECCS injection rate.

Distractor 3: Plausible because drywell radiation levels are normally proportional to reactor power. Incorrect because determining if the core is shutdown is based on control rod position and NI instrumentation.

Reference:

QCAP 0200-10 Rev 41 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 2 Question Source: Quad Cities ILT Exam Bank (QDC.03-01 NRC RO 22)

Question History: Quad 2005 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Associated objective(s):

295010.AK3.03 (CFR 41.5, 45.6)

Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE : Radiation level monitoring (RO=3.2 / SRO=3.5)

SR-0001-K02 (Freq: LIC=I)

DEFINE the standard terms and phrases used in the QGA procedures.

Page: 45 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 24 ID: QDC.ILT.16442 Points: 1.00 Given the following conditions on Unit 2:

• A small break LOCA has occurred.

• Drywell pressure is 2.6 psig and rising slowly.

• Drywell temperature is 158°F and rising slowly.

• ALL ECCS have responded as designed.

Given the above plant conditions, the Drywell Cooler fans...

A. have TRIPPED and can NOT be restarted.

B. remain RUNNING and continue to provide cooling.

C. remain RUNNING, but RBCCW must be started to restore cooling.

D. have TRIPPED, but can be restarted if the trip signal is BYPASSED.

Answer: D Answer Explanation:

With drywell pressure greater than 2.5 psig, Core Spray logic automatically load sheds the drywell cooler fans. The cooler fans can be manually started if the LOCA trip signal is automatically bypassed on the 912-5 panel using keylock switches.

Distractor 1: Plausible if candidate assumes that the drywell coolers cannot be restarted following an active load shed trip.

Distractor 2: Plausible if candidate assumes that drywell coolers are powered from non-essential busses and therefore not susceptible to a load-shed scheme.

Distractor 3: Plausible if candidate does not recognize that drywell coolers and RBCCW are load shed on a LOCA signal.

Reference:

QCOA 0201-01 Rev 23, QCOP 5750-19 Rev 7 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 2 Question Source: Hope Creek ILT Exam Bank Question History: Hope Creek 2010 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: None Comments: None Page: 46 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

295012.AK2.01 (CFR 41.7, 45.8)

Knowledge of the interrelations between HIGH DRYWELL TEMPERATURE and the following: Drywell ventilation (RO=3.4 / SRO=3.5)

SR-1602-K26 (Freq: LIC=B)

EVALUATE given key Primary Containment Atmosphere Control Systems parameter indications and/or responses depicting a system specific abnormality/failure and DETERMINE a course of action to correct or mitigate the following abnormal condition(s):

a. High air temperature

b. High/low pressure

c. High/low drywell/torus differential pressure Page: 47 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 25 ID: QDC.ILT.16443 Points: 1.00 Unit 1 operating at rated power.

Answer the following question in accordance with Technical Specification 3.6.2.1, Suppression Pool Average Temperature:

To ensure complete condensation of decay heat and sensible heat, which of the following is the LOWEST suppression pool temperature that requires placing the reactor mode switch to shutdown IMMEDIATELY?

A. 96°F B. 107°F C. 112°F D. 121°F Answer: C Answer Explanation:

The reactor must be shutdown when suppression pool temperature is > 110°F. This ensures that the suppression pool can absorb (condense) decay and sensible heat yet stay within the design limits.

Distractor 1, 2 and 3: All plausible because each one exceeds a separate Tech Spec threshold and requires separate actions.

Reference:

Tech Spec LCO 3.6.2.1 Amendment 199/195 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 2 Question Source: Duane Arnold ILT Exam Bank Question History: Duane Arnold 2007 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(8-10)

SRO Justification: N/A Comments: None Page: 48 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-1601-K28 (Freq: LIC=B)

EXPLAIN the reasons for given Containment Systems operating limits and precautions.

a. Torus temperature limits (95/110/160)

b. Torus level limits (+2/-2 adjusted for dp)

c. Drywell/torus differential pressure limitations

d. Drywell Spray Initiation Limit

e. Primary Containment Pressure Limit

f. Pressure Suppression Limit

g. Heat Capacity Limit 295013.AK1.04 (CFR 41.8 to 41.10)

Knowledge of the operational implications of the following concepts as they apply to HIGH SUPPRESSION POOL TEMPERATURE : Complete condensation (RO=2.9 /

SRO=3.2)

Page: 49 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 26 ID: QDC.ILT.16502 Points: 1.00 Unit 1 is at rated power when an INADVERTENT primary containment group (PCIS) isolation occurs.

Plant conditions allow resetting the isolation signal.

Which PCIS Group isolation(s) can be RESET using the 901-5 panel switch displayed below?

A. Group 1 only.

B. Group 2 and 3 only.

C. Group 1, 2 and 3 only.

D. All PCIS Group isolations.

Answer: B Answer Explanation:

INBD resets inboard Group 2 and 3 valves, OUTBD position resets the outboard Group 2 and 3 valves.

Distractor 1: Plausible because the Group 1 inbd/outbd isolation reset switch is next to the Group 2 & 3 isol reset switch.

Distractor 2: Plausible because Group 1, 2 and 3 isolations are reset from the 901-5 panel.

Distractor 3: Plausible if candidate assumes that all PCIS group isolations are reset using this switch (reinforced by the fact that the switch is not labeled as a specific group isolation reset switch).

Reference:

QCAN 901(2)-5 B-5 Rev 9, QCAN 901(2)-5 B-6 Rev 10 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 2 Question Source: New Page: 50 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

295020.2.1.28 Knowledge of the purpose and function of major system components and controls. (RO=4.1 / SRO=4.1)

SR-1603-K16 (Freq: LIC=I)

STATE the physical location and DESCRIBE the operation of the following Primary Containment Isolation (PCI) System controls (local/remote):

a. Main Stm Isol Valve Reset switch (Group 1)

b. Reactor mode switch (Group 1 / 2)

c. Isol Valve Reset switch (Group 2 / 3)

d. Vent Isol Signal Bypass switch (torus/drywell) (Group 2)

e. Reactor building ventilation Emergency Damper Reset switches (912-1) / (2251(2)-

24X) (Group 2)

f. Reset For Grp 2 Isol Vlv 1(2)-1001-29

g. Reactor Bldg Flr Drn Sump A/B Reset pushbuttons

h. Reactor Bldg Equip Drn Tnk Sump Reset pushbutton

i. Master vent mode switch (Norm/APCV)

j. AO 1(2)-1602-24/23/60 CIS Override keylock switches Page: 51 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 27 ID: QDC.ILT.16444 Points: 1.00 Unit 2 is at 100% power when the running CRD pump TRIPS.

A HCU accumulator TROUBLE alarm is then received for Control Rod L-12 in the Control Room.

• A red light is lit on the local CRD panel at the HCU banks. When the EO depresses the alarm light pushbutton, the light remains LIT.

Answer the following questions:

(1) What condition is CAUSING the alarm light; and (2) What is the IMPACT to Control Rod L-12 of continued plant operation in this condition?

A. (1) Low GAS PRESSURE (2) Control Rod may DRIFT into the core B. (1) Low GAS PRESSURE (2) Scram speed may be SLOWER C. (1) High WATER LEVEL (2) Scram speed may be SLOWER D. (1) High WATER LEVEL (2) Control Rod may DRIFT into the core Answer: B Answer Explanation:

If the light remains lit, the cause is low accumulator pressure.

When accumulator pressure is low (with the plant at rated power), the control rod scram speed will be slower.

Distractor 1: Plausible because other CRD system failures (e.g. FCV fails open) will result in control rods possibly drifting into the core.

Distractor 2: Plausible because this is the other input into the alarm function (light goes out if due to high water level).

Distractor 3: Combination of distractor 1 and 2.

Reference:

QCOA 0300-01 Rev 16, QCAN 901(2)-5 G-2 Rev 9, TS LCO 3.1.5 Amendment 199/195 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 2 Question Source: Modified from Quad ILT Exam Bank (QDC.LN.10418)

Question History: N/A 10 CFR Part 55 Content: 41(b)(10)

Page: 52 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

SRO Justification: N/A Comments: None Associated objective(s):

SR-0302-K06 (Freq: LIC=I)

Given a Control Rod Drive Hydraulics annunciator tile inscription or accumulator gas/water trouble light, DESCRIBE the condition causing the alarm and any automatic actions which occur when the alarm actuates. EXPLAIN the consequences of the condition if not corrected.

295022.AA2.01 (CFR 41.10, 43.5, 45.13)

Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS: Accumulator pressure (RO=3.5 / SRO=3.6)

Page: 53 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 28 ID: QDC.ILT.16456 Points: 1.00 Power ascension is in progress with Unit 1 at 10% RTP in Mode 1, when a LPRM input to APRM Channel 1 fails DOWNSCALE.

• Annunciator 901-5 E-7 "LPRM DOWNSCALE" comes in.

• Annunicator 901-5 C-6 "APRM DOWNSCALE" comes in.

• APRM Channel 1 currently has 13 input LPRMs in OPERATE.

Which one of the following identifies the impact of these conditions on continued rod withdrawal and required action?

A. A rod block exists. If the failed LPRM is bypassed, APRM Channel 1 must also be bypassed to clear the rod block.

B. A rod block exists. ONLY the failed LPRM must be bypassed to clear the rod block.

C. NO rod block exists. APRM 1 is administratively inoperable and must be bypassed.

D. NO rod block exists. The failed LPRM must be bypassed to clear the alarm.

Answer: A Answer Explanation:

A rod block is issued when any APRM is downscale with the Reactor Mode switch in RUN. As soon as the LPRM is bypassed, only 12 LPRMs will be in OPERATE which will result in another rod block (due to < 13 LPRM inputs). Therefore the APRM must be bypassed to continue rod withdrawal.

Distractor 1: Plausible if candidate does not recognize that when the LPRM is bypassed, less than the required number of LPRMs (13) will be in operate and a rod block will occur.

Distractor 2: Plausible if candidate assumes that less than 13 LPRMs results in an administratively inoperable APRM (similar to the < 2 LPRMs per level requirement).

Distractor 3: Plausible if candidate does not recognize that a rod block occurs with a APRM downscale and assumes that the APRM is operable.

Reference:

QCAN 901(2)-5 C-6 Rev 5, QCAN 901(2)-5 C-12 Rev 8, QCAN 901(2)-5 C-3 Rev 11, LIC-0703 Rev 14 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.703429)

Question History: Susquehanna 2002 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Page: 54 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Comments: None Associated objective(s):

SR-0703-K26 (Freq: LIC=B)

EVALUATE given key LPRM/APRM System parameter indications and/or responses depicting a system specific abnormality/failure and DETERMINE a course of action to correct or mitigate the following abnormal condition(s):

a. Loss of RPS power

b. Flow convertor output fails high/low

c. LPRM output fails high/low

d. APRM output fails high/low

e. Direct power indications disagree with alternate power indications 215005.A2.08 (CFR 41.5 / 45.6)

Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE 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: Faulty or erratic operation of detectors/systems (RO=3.2 / SRO=3.4)

Page: 55 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 29 ID: QDC.ILT.16458 Points: 1.00 Following an accident, RCIC is required as an injection source for makeup water to the reactor vessel.

• RCIC is manually started with the INITIATION pushbutton and is injecting into the reactor at 400 gpm.

If MO 1-1301-60, RCIC MIN FLOW VLV inadvertently OPENS, which of the following indications would the operator expect?

A. DECREASE in flow indication on the RCIC flow controller.

B. INCREASE in RCIC pump discharge pressure.

C. CCST level to decrease at a SLOWER rate than before.

D. Torus water level to increase at a FASTER rate than before.

Answer: D Answer Explanation:

Torus level will increase at a faster rate due to the opening of minimum flow valve MO 1-1301-60. The RCIC flow controller does not monitor minimum flow, therefore total RCIC system flow will be greater than 400 gpm, resulting in a faster rate of increase for Torus water level.

Distractor 1: Plausible if candidate assumes that the RCIC flow controller is in Manual when you start RCIC with the manual pushbutton.

Distractor 2: Plausible if candidate assumes that a higher system flow with the minimum flow valve open requires a higher discharge pressure of the RCIC pump.

Distractor 3: Plausible if candidate assumes that the minimum flow line returns to the CCST (similar to the test line) and the RCIC flow controller will maintain system flow at 400 gpm.

Reference:

QCOS 1300-01 Rev 37, LN-1300 Rev 10 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Cooper ILT Exam Bank Question History: Cooper 2005 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Page: 56 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

217000.A1.07 (CFR 41.5 / 45.5)

Ability to predict and/or monitor changes in parameters associated with operating the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) controls including:

Suppression pool level (RO=3.3 / SRO=3.5)

SR-1300-K21 (Freq: LIC=B)

Given a RCIC System operating mode and various plant conditions, PREDICT how RCIC/plant parameters will respond to manipulation of the following RCIC System local/remote controls:

a. Manual Initiation Pushbutton

b. RCIC Initiation Reset

c. Trip pushbutton

d. Trip Reset switch

e. RCIC Isolation Reset

f. Room cooler controls (including Hand/Auto switch)

g. FIC 1(2)-1340-1 (RCIC flow controller)

h. MOV/AOV controls (control room)

i. Local valve control stations

j. Turb Speed Test Switch

k. Bias Speed Setting potentiometer

l. Cond and vacuum pump Remote/Local switches

m. 16-valve local/remote switch

n. Cond and vacuum pump load side/source side transfer switches

o. Local Bias Speed Setting potentiometer Page: 57 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 30 ID: QDC.ILT.16459 Points: 1.00 Unit 2 is at rated power when a transient occurs. The following annunciator is IN ALARM on the 902-3 panel:

• C-15, AUTO BLOWDOWN INTERLOCK CORE SPRAY/RHR Which of the following statements CORRECTLY describes the impact of this annunciator?

A. An ADS permissive for actuation is NOT met.

B. An ADS permissive for actuation is MET because the supply BREAKER of at least one low pressure ECCS pump is closed.

C. An ADS permissive for actuation is MET because the FLOWRATE of at least one low pressure ECCS pump has exceeded the setpoint.

D. An ADS permissive for actuation is MET because the discharge PRESSURE of at least one low pressure ECCS pump has exceeded the setpoint.

Answer: D Answer Explanation:

This annunciator alarms when any one of the RHR and Core Spray pumps develop greater than 100 psig discharge pressure. This ensures that a low pressure ECCS system is able to inject into the RPV following a blowdown and is required to be met to initiate ADS.

Distractor 1: Plausible if candidate assumes that an alarming condition means that the permissive for CS/RHR is NOT met.

Distractor 2: Plausible because there is another annunciator that alarms when a RHR or CS pump breaker is closed.

Distractor 3: Plausible because both the RHR and CS pumps will be running on minimum flow, therefore able to inject into the RPV following a blowdown.

Reference:

QCAN 901(2)-3 C-15 Rev 7 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 58 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

218000.K6.01 (CFR 41.7 / 45.7)

Knowledge of the effect that a loss or malfunction of the following will have on the AUTOMATIC DEPRESSURIZATION SYSTEM : RHR/LPCI system pressure: Plant-Specific (RO=3.9 / SRO=4.1)

SR-0203-K06 (Freq: LIC=I)

Given an ADS logic/valve annunciator tile inscription, DESCRIBE the condition causing the alarm and any automatic actions which occur when the alarm actuates. EXPLAIN the consequences of the condition if not corrected.

Page: 59 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 31 ID: QDC.ILT.16461 Points: 1.00 Unit 2 is at rated power when Bus 29 TRIPS.

Which of the following CORRECTLY describes the plant response to this event?

A. 2B Feedwater Regulating Valve (FRV) locks-up B. Reactor Water Cleanup (RWCU) inboard and outboard suction valves close C. Steam Jet Air Ejector (SJAE) suction valves close D. Intermediate Range Monitor (IRM) Channel 15 goes downscale.

Answer: B Answer Explanation:

A loss of Bus 29 will result in a loss of RPS B (120 VAC). A loss of RPS B will result in a 1/2 Group 3 isolation. Unlike a Group 1 isolation, a 1/2 Group 3 isolation is enough to close the RWCU PCI valves.

Distractor 1: Plausible because a loss of the 120 VAC Instrument Bus will lock-up the 2B FRV.

Distractor 2: Plausible because a loss of the 120 VAC Essential Service Bus will result in a closure of the SJAE suction valves.

Distractor 3: Plausible because APRMs 4-6 lose power. Incorrect because IRMs are powered from 24/48 VDC.

Reference:

QOA 7000-01 Rev 33, QOA 6700-04 Rev 23 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Duane Arnold ILT Exam Bank Question History: Duane Arnold 2002 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 60 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

223002.K6.08 (CFR 41.7 / 45.7)

Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF : Reactor protection system (RO=3.5 / SRO=3.7)

SR-1603-K22 (Freq: LIC=B)

Given various plant conditions, PREDICT how Primary Containment Isolation (PCI)

Systems and key plant parameters will respond to the following failures:

a. Loss of RPS A and/or B power

b. Loss of Essential Service bus

c. Loss of 125vdc

d. Loss of air/drywell pneumatics

e. Loss of Instrument Bus Page: 61 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 32 ID: QDC.ILT.16462 Points: 1.00 Unit 1 was at full power when a transient caused a reactor scram on high reactor pressure.

• RPV pressure is currently being controlled by the automatic cycling of relief valves.

What effect does the "INHIBIT" light of the 1-203-3B and 3C relief valves being ILLUMINATED have on the current pressure control method?

An illuminated INHIBIT light...

A. is an indication ONLY and will NOT prevent the relief valves from opening.

B. indicates that the relief valves are PREVENTED from being opened MANUALLY and from opening on HIGH reactor pressure.

C. indicates that the relief valves are PREVENTED from opening on HIGH reactor pressure, but NOT prevented from being opened MANUALLY.

D. indicates that the relief valves are PREVENTED from being opened MANUALLY, but NOT prevented from opening on HIGH reactor pressure.

Answer: C Answer Explanation:

Relief Valves 3B and 3C are interlocked to prevent automatic valve reopening within 10 seconds after valve closing. Actuation of this interlock is indicated by illumination of the amber INHIBIT LIGHT. This interlock does NOT prevent manual opening of the valve but manual operation within this 10 second period is prohibited by procedure. To prevent reopening of the Relief Valves within 10 seconds following valve closure, the interlock timers have been set for approximately 14.5 seconds to account for potential delays in valve closure.

Distractor 1: Plausible if candidate assumes that it is indication only to allow the operator to prevent excessive torus loading during operation.

Distractor 2 and 3: Plausible if candidate incorrectly applies the design of the interlock.

Reference:

QCOP 0203-01 Rev 13 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.00707)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 62 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

239002.K4.03 (CFR 41.7)

Knowledge of RELIEF/SAFETY VALVES design feature(s) and/or interlocks which provide for the following: Prevents siphoning of water into SRV discharge piping and limits loads on subsequent actuation of SRV's (RO=3.1 / SRO=3.3)

SR-0203-K13 (Freq: LIC=I)

DESCRIBE the 14.5 second interlock associated with the 'B' and 'C' electromatic relief valves/ PORVs, including purpose and when/how it is bypassed.

Page: 63 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 33 ID: QDC.ILT.16463 Points: 1.00 Given the following conditions:

• Unit 2 is at 50% power

• DFWLC is in 3-element

• Both FRVs and the LFFRV are in AUTO

• The 2A and 2B RFPs are running and 2C RFP is selected for standby If the feedwater flow transmitter for the 2A RFP slowly fails DOWNSCALE (soft failure), what effect will this have on reactor water level?

A. Reactor water level INCREASES momentarily due to the reduced feedflow signal, then RETURNS to +30 inches.

B. Reactor water level INCREASES continuously due to the reduced feedflow signal until the RFP TRIP setpoint is reached.

C. Reactor water level remains CONSTANT throughout the transient because the 2A RFP signal is removed as an input to the DFWLC system.

D. Reactor water level DECREASES momentarily, then RETURNS to +30 inches when the 2C RFP auto-starts.

Answer: A Answer Explanation:

When a Feed Flow transmitter fails low but is not perceived by the system as a signal error (i.e. signal still in the 4-20 mA range), the system remains in 3-Element Control.

Feed demand based upon the Level Controller and Total Steam Flow inputs remains the same. The Feedwater Controller compares this demand with measured flow and finds measured flow to be lacking since one of its feed flow transmitters has failed low. Thus the Feedwater Flow Controller increases its output and causes the FRVs to open.

This causes RPV Level to increase. As a deviation develops between actual level and the Level Setpoint, the Level Controller turns level and causes a subsequent reduction in actual FW Flow and restores level to the Level Setpoint. [Remember, the system should be able to deal with a flow measurement error of up to 5 Mlbm/hr and allow the Level Controller to restore level to the setpoint value.]

At 50% power, each RFP will have a flow of approximately 3 Mlb/hr.

Distractor 1: Plausible if candidate does not recognize that the level inputs to 3-element control are the dominating factor.

Distractor 2: Plausible because a soft majority selector (SMS) switch is available to remove reactor water level inputs. Incorrect because DFWLC does not use a SMS for feedwater flow.

Distractor 3: Plausible because the 2A RFP has auto-start features. However, the feed breaker to a running RFP must trip to auto-start the standby pump.

Reference:

QCOP 0600-21 Rev 16, LIC-0600 Rev 8 Reference provided during examination: None Page: 64 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.L.631820)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

SR-0600-K24 (Freq: LIC=B Given a Feedwater Level Control System operating mode and various plant conditions, PREDICT how each supported system will be impacted by the following Feedwater Level Control System failures:

a. Feedflow sensor failures

b. Steam flow sensor failures

c. RPV level sensor failures

d. Reactor level SMS value error

e. RFP suction pressure SMS value error

f. Steam flow SMS value error

g. Feedflow error with calculated feedwater flow activated 259002.K3.01 (CFR 41.7 / 45.5 to 45.8)

Knowledge of the effect that a loss or malfunction of the REACTOR WATER LEVEL CONTROL SYSTEM will have on following: Reactor water level (RO=3.8 / SRO=3.8)

Page: 65 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 34 ID: QDC.ILT.16465 Points: 1.00 (Refer to the FOUR attached pictures of the Standby Gas Treatment System (SBGTS) mimic to answer this question.)

Which lineup shows the CORRECT position for the SBGTS motor operated valve dampers if both trains are in their normal standby lineup?

A. Lineup A B. Lineup B C. Lineup C D. Lineup D Answer: C Answer Explanation:

Lineup C depicts the normal standby lineup for both trains of SBGT:

1(2)-7503 UR RB INLET DMPR TO SBGTS is OPEN 1/2-7505A/B INLET DMPR is CLOSED 1/2-7504A/B TURB BLDG CLG AIR DMPR is OPEN 1/2-7507A/B SBGTS FAN DISCH DMPR is CLOSED 1/2-7509 XTIE DMPR is OPEN (Note: the 'B' Train of SBGTS is normally in Primary.)

Lineup A: Show picture with all SBGTS dampers closed.

Lineup B: Show picture with all SBGTS 'A' train dampers closed.

Lineup D: Show picture with all SBGTS train dampers open except the 7507A/B damper and the 7509 damper.

Distractor 1: Plausible if candidate assumes that ALL dampers are closed when SBGTS is not in operation.

Distractor 2: Plausible if candidate assumes that all 'A' train dampers are closed because the SBGTS selector switch is normally in PRIMARY for the 'B' train.

Distractor 3: Plausible if candidate assumes that all of the dampers, except the discharge dampers and crosstie damper, are open when in standby.

Reference:

QCOP 7500-01 Rev 19 Reference provided during examination: None Cognitive level: Low Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

Page: 66 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

SRO Justification: N/A Comments: None Associated objective(s):

SR-7500-K20 (Freq: LIC=B)

Given a SBGTS operating mode and various plant conditions, EVALUATE the following SBGTS indications/responses and DETERMINE if the indication/ response is expected and normal.

a. System flow

b. Fan and heater indicating lights

c. MOV position

d. Air-operated outlet valve 1/2-7510A/B position

e. Heater inlet and outlet temperatures

f. Heater differential temperature

g. Run-time

h. Train component differential pressures

i. Charcoal adsorber temperature 261000.A4.09 (CFR 41.7 / 45.5 to 45.8)

Ability to manually operate and/or monitor in the control room: Ventilation valves/dampers (RO=2.7 / SRO=2.7)

Page: 67 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 35 ID: QDC.ILT.16467 Points: 1.00 The Unit 1 Essential Service Bus transferred from its NORMAL power supply to one of the backup power supplies, during which a MOMENTARY loss of Essential Service occurred.

This type of transfer is indicative of an Essential Bus transfer to which of the following power supplies?

A. Bus 18 B. Bus 17 C. MCC 18-2 D. 250 VDC MCC-1 Answer: C Answer Explanation:

MCC 18-2 supplies backup power to the Essential Bus through an ABT, but this results in momentary loss of power. Other ESS Bus transfers to the DC supply or to the regulated AC supply (from SWGR 17 through the Static Switch) would not interrupt ESS Bus Power.

Distractor 1, 2 and 3: All three alternate answer options can provide power to the ESS Bus but automatic transfers to them do not interrupt ESS Bus power.

Reference:

QOA 6800-03 Rev 41, LN-6800 Rev 13 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.627437)

Question History: N/A 10 CFR Part 55 Content: 41(b)(2-9)

SRO Justification: N/A Comments: None Page: 68 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SRN-6800-K20 (Freq: LIC=B NF=B)

Given an Essential Service/Instrument Bus Systems operating mode and various plant conditions, EVALUATE the following Essential Service/Instrument Bus Systems indications/responses and DETERMINE if the indication/ response is expected and normal.

a. ESS UPS (1) Battery input amps/volts (2) Normal DC rectifier output amps/volts (3) Inverter AC output volts/frequency (4) Alternate AC regulator output volts (5) AC load amps/volts (6) Status lights

b. Instrument bus voltage

c. Aux electric room ambient temperature 262001.K1.04 (CFR 41.2 to 41.9 / 45.7 to 45.8)

Knowledge of the physical connections and/or cause- effect relationships between A.C.

ELECTRICAL DISTRIBUTION and the following: Uninterruptible power supply (RO=3.1

/ SRO=3.4)

Page: 69 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 36 ID: QDC.ILT.16504 Points: 1.00 Unit 2 is at 100% power when the following annunciator is received on the 902-8 panel:

• G-9, 120V AC INSTR BUS TRANSFER TO EMERG SPLY Complete the following statements:

Based on this annunciator, an operator must verify at panel 902-50 that the auto-transfer switch (ABT) is CLOSED onto __(1)__.

When NORMAL power is restored, the operating crew must __(2)__ to its normal power supply.

A. (1) MCC 25-2 (2) verify that the Instrument Bus AUTOMATICALLY transfers B. (1) MCC 25-2 (2) MANUALLY transfer the Instrument Bus C. (1) MCC 28-2 (2) verify that the Instrument Bus AUTOMATICALLY transfers D. (1) MCC 28-2 (2) MANUALLY transfer the Instrument Bus Answer: A Answer Explanation:

MCC 28-2 is the normal supply through the Instrument Bus ABT. The ABT automatically swaps power to MCC 25-2 on loss of the normal supply.

There are no controls that are operated in response to this alarm. The ABT is "normal-seeking" and will transfer back to MCC 28-2 when normal power is restored.

Distractor 1: Plausible because the Essential Service Bus ABT is "power-seeking".

Distractor 2: Plausible because MCC 28-2 is the normal supply through the Instrument Bus ABT.

Distractor 3: Combination of distractor 1 and 2.

Reference:

QOA 900-8 G-9 Rev 4, QCOP 6800-02 Rev 16 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 2 Question Source: Modified from Quad Cities ILT Exam Bank (QDC.ILT.627455)

Question History: N/A 10 CFR Part 55 Content: 41(b)(10)

SRO Justification: N/A Page: 70 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Comments: None Associated objective(s):

SR-6800-K06 (Freq: LIC=I)

Given an Essential Service/Instrument Bus Systems remote annunciator tile inscription or ESS UPS local alarm light, DESCRIBE the condition causing the alarm and any automatic actions which occur when the alarm actuates. EXPLAIN the consequences of the condition if not corrected.

262002.2.4.50 Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. (CFR: 45.3) (RO=4.2 / SRO=4.0)

Page: 71 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 37 ID: QDC.ILT.16468 Points: 1.00 Complete the following statements regarding the operation of the 125 VDC System Ground Detector:

The presence of a LEVEL III ground on the 125 VDC system __(1)__ annunciate an alarm in the Control Room.

When cycling 125 VDC breakers, GROUND DETECTOR VOLTAGE will indicate __(2)__ when all grounds have been ISOLATED and the positive and negative-to-ground pushbuttons are NOT being depressed.

A. (1) will (2) zero (0) VDC B. (1) will NOT (2) zero (0) VDC C. (1) will (2) 125 VDC (positive or negative)

D. (1) will NOT (2) 125 VDC (positive or negative)

Answer: A Answer Explanation:

The ground detector is aligned such that with neither pushbutton depressed and no grounds present, current flow through the calibrated resistors will be equal in magnitude and opposite in polarity such that the 2 currents being measured will cancel each other out. The net current flow resulting will be 0 and the calibrated voltmeter will read 0 volts.

A ground on either bus will cause an unbalance in the voltage drop across the resistors.

This results in a voltage felt across the ammeter to ground, current through the ammeter, and a meter deflection proportional to the magnitude of the ground.

Distractor 1: Plausible if candidate assumes that a Level III ground is the smallest ground level. Incorrect because a Level I ground (which will not annunciate the alarm) is the smallest.

Distractor 2: Plausible because the reading will be + or - when you depress the pushbuttons to check for grounds.

Distractor 3: Combination of distractor 1 and 2.

Reference:

QCOP 6900-19 Rev 11, LN-6900 Rev 16 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.688595)

Question History: N/A Page: 72 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

SRN-6900-K15 (Freq: LIC=B N=B)

DESCRIBE the operation of the following principal Station DC Electrical Systems components:

a. 250 VDC system (1) Battery chargers (a) Float (b) Equalize (2) 250 VDC breakers (3) RB MCC 1A/1B (2A/2B) copper links (including metal finger adjustment)

(4) Ground detector (calculate grounds)

b. Non-Essential 250 VDC system (1) Battery chargers (2) Breakers (3) Ground detector

c. 125 VDC system (1) Battery chargers (2) Alternate battery chargers (3) Fusible switches (4) Breakers (5) RB distribution panel 1 (2) copper links (6) Ground detector (calculate grounds)

d. 24/48 VDC system (1) Battery chargers (2) Breakers

e. Fluke (1) DC bus voltages 263000.A4.04 (CFR 41.7 / 45.5 to 45.8)

Ability to manually operate and/or monitor in the control room: Ground detection circuit:

Plant-Specific (RO=3.0 / SRO=3.2)

Page: 73 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 38 ID: QDC.ILT.16469 Points: 1.00 Unit 2 was operating at rated power when a transient occurred. The following conditions are present:

• Loss of Off-site Power

• Reactor water level is -30 inches and rising

• Drywell pressure is 4.5 psig and rising Which of the following sets of U2 EDG indications would the NSO expect to see on the 902-8 Panel given the above plant conditions?

A.

B.

C.

D.

Page: 74 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Answer: B Answer Explanation:

With Bus 24-1 undervoltage (due to the LOOP), the candidate must know that the Unit 2 EDG will auto-start and energize Bus 24-1. The acceptance criteria for a successful "Fast Start" is: 3952 to 4368 Volts AND 58.8 to 61.2 Hz within <13 seconds.

The candidate must also know that the EDG governor is in isochronous when in standby, therefore maintaining voltage and frequency at the nominal setpoint independent of loading.

Distractor 1: Plausible if candidate assumes that the governor operates with a speed droop (similar to surveillance operations).

Distractor 2: Combination of distractor 1 and 3.

Distractor 3: Plausible if candidate assumes that the voltage regulator operates with a voltage droop (similar to surveillance operations).

Reference:

QCOA 6100-03 Rev 26, TS LCO 3.8.1 Amendment 245/240 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.699435)

Question History: Dresden 2009 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 75 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SRN-6600-K20 (Freq: LIC=B NF=B) Given an Emergency Diesel Generators operating mode and various plant conditions, EVALUATE the following local Emergency Diesel Generators indications/responses and DETERMINE if the indication/ response is expected and normal.

a. DG Room Ventilation (1) Vent fan status lights (2) Vent fan norm/alt feed indicating lights (3) Vent fan fire protection bypass status lights (4) U1/2 vent fan power supply indications (5) DG vent damper N2 supply pressure (6) DG room temperature

b. Diesel Cooling Water (1) DGCWP norm/alt feed indicating lights (2) DGCWP status lights (3) DGCWP oil level (4) DGCWP flow and suction/discharge pressure (5) Cooling water temperature/pressure (Heat exchanger inlet and outlet)

(6) Flow (DG heat exchanger / ECCS room coolers (RHR/CS-RCIC/HPCI))

(7) Cooling water inlet and outlet pressures (DGCWP cubicle cooler / ECCS room coolers)

c. Diesel Closed Cooling Water (1) Expansion tank level (2) DG engine driven cooling water pump pressure (3) Engine inlet and outlet temperatures

d. Diesel Lube Oil (1) Crankcase, lube oil cooler tank, and engine oil levels (2) Temperatures

e. Diesel Air Start (1) Starting air compressor status lights (2) Starting air receiver pressures (3) Air start pressure after PRV

g. Diesel Fuel Oil (1) Fuel oil storage tank levels (a) DG fuel tank level indicator air supply flow rate 1(2)-5241-20 / 1/2 5241-21 (2) Fuel oil day tank levels (3) Fuel oil pressure (4) Fuel oil transfer pump discharge pressure (5) Fuel oil transfer pump status lights (6) Fuel oil transfer pump norm/alt feed indicating lights (7) Fuel rack position

h. Diesel engine (1) Engine speed (2) Cylinder exhaust temperatures (3) Governor oil level

i. Generator (1) Volts, amps, frequency, kilowatts, watt-hours, field current, VARS

j. 1/2 EDG Aux Services Transfer Logic Power Available light 264000.A3.01 (CFR 41.7 / 45.7)

Ability to monitor automatic operations of the EMERGENCY GENERATORS (DIESEL/JET) including: Automatic starting of compressor and emergency generator (RO=3.0 / SRO=3.1)

Page: 76 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 39 ID: QDC.ILT.16508 Points: 1.00 Unit 1 is shutdown for a maintenance outage with the primary containment DE-INERTED.

• The Drywell pneumatic suction filter becomes clogged from foreign material.

Complete the following statements:

With the suction filter clogged, the Drywell pneumatic air compressor will TRIP on __(1)__.

If AO 1-4723, RES SPLY, fails CLOSED, an operator must manually open 1-4799-168, BACKUP SPLY REG BYPASS VLV to valve in the __(2)__ system.

A. (1) low suction pressure (2) Nitrogen Make-up B. (1) high air outlet temperature (2) Nitrogen Make-up C. (1) low suction pressure (2) Instrument Air D. (1) high air outlet temperature (2) Instrument Air Answer: C Answer Explanation:

With the suction filter clogged, vacuum will increase in the suction line of the air compressor until the compressor trips on low suction pressure. This will result in low drywell pneumatic pressure.

Per the annunciator response procedure, the 1-4799-168 is opened if the normal supply valve fails to automatically open. With the containment de-inerted, Instrument Air is aligned as the back-up source.

The second part of the question is high cognitive because the candidate must recognize that the primary containment conditions (de-inerted) require that the Instrument Air system is used as the backup system. The Nitrogen Make-up system is the normal backup supply when the primary containment is inerted.

Distractor 1: Plausible because Nitrogen Make-up is normally the backup supply when operating at rated conditions.

Distractor 2: Combination of distractor 1 and 3.

Distractor 3: Plausible if candidate assumes that there is not a low suction pressure trip (similar to Instrument Air compressors) and assumes that continued operation without a source will result in a high air outlet temperature. Incorrect because there is not a trip on high air outlet temperature for the drywell pneumatic air compressor like there is on the Instrument Air compressors.

Reference:

QOP 4700-06 Rev 9, QOA 900-4 G-13 Rev 3, LN-4702 Rev 4 Reference provided during examination: None Page: 77 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Associated objective(s):

SRN-4702-K23 (Freq: LIC=B NF=B) Given a Drywell Pneumatics System operating mode and various plant conditions, PREDICT how the Drywell Pneumatics System will be impacted by the following support system failures:

a. Loss of RBCCW

b. Loss of clean demin

c. Loss of 480 vac 300000.A2.01 (CFR 41.5 / 45.6)

Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Air dryer and filter malfunctions (RO=2.9 /

SRO=2.8)

Page: 78 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 40 ID: QDC.ILT.16506 Points: 1.00 Unit 1 is at rated power when a Service Water leak occurs.

• The NSO places the Unit 1 Service Water Header Isolation Switch on the 912-1 panel to CLOSE.

Assuming the NSO's actions successfully ISOLATE the leak, which of the following indications is EXPECTED due to the positioning of the header isolation switch?

A. MSIV Room temperature RISES B. TBCCW temperature RISES C. Fire Main Header pressure LOWERS D. Turbine Lube Oil temperature RISES Answer: B Answer Explanation:

Placing the Unit 1 Service Water header isolation valve to CLOSE will shut the following valves:

MO 1-3903A, TBCCW HX MO 1-3903B, TBCCW HX MO 1-3904, U1 ASD COOLERS SW DISCH VLV MO 1-3905, U1 STATOR SW SUPPLY With Service Water isolated to the TBCCW heat exchangers, TBCCW temperature will rise.

Distractor 1, 2 and 3: All distractors are plausible based on the candidate's misconception of which loads are isolated and which loads are NOT isolated by the manipulation of the header switch.

Reference:

QCOA 3900-01 Rev 16, LN-3900 Rev 3 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Modified from Quad Cities Bank (QDC.LORTB.12481)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 79 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

400000.A1.02 (CFR 41.7 / 45.7)

Ability to predict and / or monitor changes in parameters associated with operating the CCWS controls including: CCW temperature (RO=2.8 / SRO=2.8)

SR-3900-K21 (Freq: LIC=B)

Given a Service Water System operating mode and various plant conditions, PREDICT how service water system/plant parameters will respond to manipulation of the following Service Water System local/remote controls:

a. Pump control switch

b. SW header isolation switch (3905, 3904 1A/1B-3903, 2A/2B-3903)

c. Standby Coolant Supply switches (3901, 3902)

d. Service water supply to firemain valve MO-3906

e. Strainer control panel power switch

f. Strainer mode selector switches

g. Strainer basket selector switches

h. Strainer auto backwash start pushbutton Page: 80 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 41 ID: QDC.ILT.16471 Points: 1.00 Unit 1 and Unit 2 are at rated power.

• SSMP system is in its NORMAL lineup.

Which of the following busses, if de-energized, will REMOVE power from MOV 1-2901-8, U1 REACTOR SUPPLY VLV?

A. MCC 2 B. MCC 19-1 C. Bus 14-1 D. Bus 24-1 Answer: C Answer Explanation:

Under normal plant conditions, MCC 30 is energized from Bus 14-1. MCC 30 supplies 480 VAC power to the Unit 1 SSMP injection valve (1-2901-8).

Distractor 1: Plausible because 250 VDC MCC 2 supplies power to HPCI and RCIC injection valves.

Distractor 2: Plausible because MCC 19-1 supplies power to the Core Spray injection valve.

Distractor 3: Plausible because Bus 24-1 is the alternate power supply to MCC 30.

Reference:

QOM 1/2-6700-T05 Rev 3 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 81 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-2900-K19 (Freq: LIC=B)

LIST the plant systems which support the SSMP system and DESCRIBE the nature of support. (Includes power supplies)

SSMP K2.01 (CFR 41.7)

Knowledge of electrical power supplies to the following: Motor operated valves (RO=2.8

/ SRO=2.8)

Page: 82 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 42 ID: QDC.ILT.16446 Points: 1.00 An Anticipated Transient Without Scram (ATWS) blowdown is in progress on Unit 2.

• Reactor pressure is 95 psig and lowering.

• Off-site power is NOT available and the Main Turbine is tripped.

• QGA actions require injecting into the reactor to raise water level.

Given the current plant conditions, which of the following is the PREFERRED injection source for raising reactor water level, and why?

A. Core Spray (CS) because spray cooling is the preferred method of heat removal from the core.

B. Low Pressure Coolant Injection (LPCI) because the system injects water outside of the core shroud.

C. High Pressure Coolant Injection (HPCI) because the system injects high quality water from the CCSTs.

D. Condensate & Feedwater because the system can quickly raise reactor water level to its normal operating band.

Answer: B Answer Explanation:

Following an ATWS blowdown, reactor water level must be raised to establish core cooling. LPCI is listed under Detail G as a preferred ATWS system because it is easy to align and injects outside of the core shroud. This allows the cold LPCI injection water to mix with the hot water before entering the core, thereby minimizing the risk of a power excursion.

Distractor 1: Plausible because spray cooling is the next preferred method of decay heat removal following core submergence. Incorrect because spraying inside of the shroud increases the risk of a power excursion.

Distractor 2: Plausible because CCST water has a higher quality than Torus water.

Incorrect because HPCI isolates at 100 psig reactor pressure.

Distractor 3: Plausible because Condensate & Feedwater is a preferred ATWS system.

Incorrect because the system is not available following a loss of off-site power and you also do not want to raise water level quickly during an ATWS.

Reference:

QGA 101 Rev 13 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(5)

Page: 83 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

SRO Justification: N/A Comments: None Associated objective(s):

SR-0001-K60 (Freq: LIC=B)

Given QGA 101, 'RPV Control (ATWS)', EXPLAIN the reasons for the limits, cautions, notes and parameter control bands.

203000.K5.02 (CFR 41.5 / 45.3)

Knowledge of the operational implications of the following concepts as they apply to RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) : Core cooling methods (RO=3.5 /

SRO=3.7)

Page: 84 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 43 ID: QDC.ILT.16447 Points: 1.00 Unit 1 is in Mode 3 with the 'A' loop of Shutdown Cooling in operation.

• The 1A RHR pump is running.

Which of the following conditions will result in an automatic TRIP of the 1A RHR pump?

Condition 1: MO 1-1001-43A, 1A SDC OR FUEL POOL SUCT VLV is closed.

Condition 2: MO 1-1001-28A, A LPCI LOOP UPSTREAM SV is throttled fully closed.

Condition 3: FUEL POOL CLG SUCTION PERMISSIVE switch is placed in OVERRIDE and MO 1-1001-50, SDC INBD ISOL VLV is closed.

(Note: Consider each condition separately.)

A. Condition 1 only.

B. Conditions 1 and 2 only.

C. Conditions 1 and 3 only.

D. Conditions 1, 2 and 3.

Answer: A Answer Explanation:

When in the Shutdown cooling mode of operation, a running RHR pump will trip (or cannot be started) unless the following valve lineup is established:

MO 1-1001-43A is fully open AND MO 1-1001-47 AND MO 1-1001-5 are fully open This interlock is to prevent pump damage from a loss of NPSH.

Distractor 1: Plausible if candidate assumes that low system flow will trip the running RHR pump (similar to SSMP system).

Distractor 2: Plausible if candidate does not recognize that placing the Fuel Pool Cooling Permissive switch in override disables the NPSH interlocks for the 47 and 50 valves.

Distractor 3: Combination of distractor 1 and 2.

Reference:

QCOA 1000-03 Rev 8 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Modified from Grand Gulf ILT Exam Bank Question History: Grand Gulf 2005 ILT NRC Exam Page: 85 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

SR-1000-K13 (Freq: LIC=B)

DESCRIBE the interlocks associated with the following RHR and RHRSW components, including purpose, setpoints, and when/how they are bypassed, including jumpers.

a. RHR inboard/outboard injection valves (MO 1001-28A/B and 29A/B)

b. RHR heat exchanger bypass (MO 1001-16A/B)

c. RHR containment cooling valves (MO 1001-23,26,34,36,37)

d. RHR pump start (suction valves/power source/FPC)

e. RHRSW pump/valves

f. RHR MO 1001-7A/B/C/D and MO 1001-43A/B/C/D

g. RHR MO 1001-19A/B, MO 1001-43A/B/C/D, and MO 1001-34A/B 205000.K4.04 (CFR 41.7)

Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following: Adequate pump NPSH (RO=2.6 / SRO=2.6)

Page: 86 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 44 ID: QDC.ILT.16501 Points: 1.00 Unit 1 was at rated power when a small LOCA with a loss of Offsite power (LOOP) occurred.

• HPCI auto-started and is injecting into the RPV at 5000 gpm.

• RPV water level is stable at +30 inches.

If the flow input to the HPCI flow controller fails high (i.e., senses higher than actual flow),

complete the following statement describing the HPCI system response and effect on RPV water level.

With NO operator action, RPV water level will...

A. LOWER because the HPCI turbine will trip on overspeed.

B. REMAIN stable because the MSC will MAINTAIN turbine speed constant.

C. REMAIN stable because the governor will MAINTAIN turbine speed constant.

D. LOWER because the MGU will REDUCE turbine speed until HPCI is no longer injecting.

Answer: D Answer Explanation:

With the flow controller in AUTO (normal position for an auto-start), the flow controller failing high will cause the motor gear unit to lower speed (sensed flow is greater than actual flow). HPCI takes the lower of the motor speed changer (variable), motor gear unit (variable), or governor (static at 4000 rpm). Because the motor gear unit (MGU) would be the lowest rpm demand, the turbine speed is reduced.

Distractor 1: Plausible if assumed all control components failed.

Distractor 2: Plausible if it is assumed the MSC controls speed, not the MGU in this case.

Distractor 3: Plausible if assumed the governor controls turbine speed, not the MGU in this case.

Reference:

LN-2300 Rev 17 Reference provided during examination: No Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Bank (QDC.L.503433)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 87 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-2300-K22 (Freq: LIC=B)

Given a HPCI System operating mode and various plant conditions, PREDICT how key HPCI system/plant parameters will respond to the following HPCI system component or controller failures:

a. Flow controller failure high/low

b. Exhaust vacuum breaker failure

c. Turning gear failure to engage

d. Exhaust rupture diaphram failure

e. Backleakage in HPCI pump discharge 206000.K3.01 Knowledge of the effect that a loss or malfunction of the HIGH PRESSURE COOLANT INJECTION SYSTEM will have on following: (CFR 41.7 / 45.4)

Reactor water level control: BWR-2,3,4 (RO=4.0 / SRO=4.0)

Page: 88 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 45 ID: QDC.ILT.16448 Points: 1.00 Unit 1 is at rated power when the following annunciator ALARMS at the 901-3 panel:

• C-5, CORE SPRAY SYS 1 BUS/LOGIC PWR FAILURE Assuming all Unit 1 4KV busses are energized, this annunciator indicates that a LOSS of which of the following has occurred?

A. 24/48 VDC Bus 1A B. Instrument Bus C. Turbine Building Main Bus 1A-1 D. Reactor Building Distribution Panel 1 Answer: C Answer Explanation:

Division 1 of Core Spray logic is powered from TB Main Bus 1A-1.

Distractor 1: Plausible because 24/48 VDC is a DC system used for instrumentation.

Distractor 2: Plausible because the Division 1 Analog Trip System uses power from the Instrument Bus.

Distractor 3: Plausible because Reactor Building Distribution Panel 1 is the normal control power feed to Bus 13-1. If the normal feed is lost, an autotransfer switch swaps the supply to 125 VDC SBO Distribution panel 6A-1 preventing an undervoltage trip of Bus 13-1. This condition is alarmed on the 901-74 (SBO) panel.

Reference:

QCOA 1400-02 Rev 10 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Quad ILT Exam Bank (QDC.ILT.627515)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 89 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

209001.K2.03 (CFR 41.7)

Knowledge of electrical power supplies to the following: Initiation logic (RO=2.9 /

SRO=3.1)

SR-1400-K19 (Freq: LIC=I)

LIST the plant systems which support the following systems and DESCRIBE the nature of support (Include power supplies):

a. Core Spray System

b. ECCS Keep Fill System Page: 90 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 46 ID: QDC.ILT.16498 Points: 1.00 What effect, if any, does a loss of Service Air have on the Standby Liquid Control (SBLC)

System?

A. No effect.

B. Loss of storage tank level indication only.

C. Loss of storage tank mixing capability only.

D. Loss of storage tank level indication and mixing capability.

Answer: C Answer Explanation:

Service air supplies the storage tank sparger used for mixing the solution or prior to sample being taken. A pressure control valve (PCV 1(2)-4601) is provided locally to adjust service air pressure to the value called for in the sparging procedure. The PCV is located on the 666 6 level near the tank.

Distractor 1: Plausible if candidate assumes that other plant air systems support the mixing of the SBLC tank.

Distractor 2: Combination of distractor 1 and 3.

Distractor 3: Plausible because the tank level indication uses Instrument Air to function.

Reference:

LIC-1100 Rev 17 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.627485)

Question History: N/A 10 CFR Part 55 Content: 41(b)(2-9)

SRO Justification: N/A Comments: None Page: 91 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

211000.K1.03 (CFR 41.2 to 41.9 / 45.7 to 45.8)

Knowledge of the physical connections and/or cause- effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following: Plant air systems: Plant-Specific (RO=2.5 / SRO=2.6)

SR-1100-K23 (Freq: LIC=B)

Given a SBLC operating mode and various plant conditions, PREDICT how the SBLC will be impacted by the following support system failures: (Includes power supplies)

a. Storage Tank Heater Power Supply

b. Pump Suction Piping Heater Power Supply

c. Pump/Squib Valve Power Supply

d. Instrument Air

e. Service Air

f. Clean Demin Page: 92 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 47 ID: QDC.ILT.16452 Points: 1.00 (Use your provided reference to answer this question.)

Given the following:

• Unit 1 was at 100% power when a transient occurred.

• Both manual Reactor scram pushbuttons have been depressed.

• The Reactor Mode switch was placed in SHUTDOWN.

• ARI was INITIATED.

• All Scram Solenoid Group lights are LIT.

• All Individual Blue Scram lights are OUT.

You are assigned to execute QCOP 0300-28, Alternate Control Rod Insertion.

Which of the following actions is the PREFERRED method for Control Rod insertion and why?

A. Perform single Control Rod scram insertion because it is the fastest method to start reducing Reactor power.

B. Bypass the RWM and manually insert CRAM Rods because it is the fastest method to start reducing Reactor power.

C. Bypass Reactor scram signals and ARI, reset and then insert a manual scram because it is a method for inserting all control rods simultaneously.

D. De-energize scram solenoids by pulling fuses at panels 901-15 and 901-17 because it is a method for inserting control rods simultaneously.

Answer: D Answer Explanation:

Scram solenoid group lights being lit indicates that RPS has not processed a scram. The individual blue scram lights being out indicate that the scram inlet and outlet valves are still shut (ARI and RPS have not worked). Per QCOP 0300-28, pulling fuses will de-energize the scram pilot solenoid valves and allow the control rods to scram. This is the preferred method of shutting down the Reactor because it provides simultaneous insertion of all control rods.

Distractor 1: Plausible if candidate assumes indications are of a hydraulic ATWS, in which single control rod insertion would be appropriate.

Distractor 2: Although an appropriate action during an electric ATWS, it is not the preferred method of inserting control rods because it inserts them individually.

Distractor 3: Plausible if candidate assumes indications are of a hydraulic ATWS, in which bypassing and resetting the RPS scram will be appropriate.

Reference:

QCOP 0300-28 Rev 29 Reference provided during examination: QCOP 0300-28 Rev 29 Attachment A only Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Page: 93 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Question Source: Quad Cities ILT Exam Bank (QDC.ILT.15483)

Question History: Quad Cities 2009 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(10)

SRO Justification: N/A Comments: None Associated objective(s):

SR-0001-K61 (Freq: LIC=B)

Given QGA 101, 'RPV Control (ATWS)', EXPLAIN the reasons for the actions.

212000.2.4.06 Knowledge of EOP mitigation strategies. (RO=3.7 / SRO=4.7)

Page: 94 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 48 ID: QDC.ILT.16453 Points: 1.00 A plant startup is in progress on Unit 1 with the Reactor Mode Select switch in STARTUP / HOT STANDBY position.

• ALL IRM Channels are on Range 5.

(Note: Reference the picture of the IRM recorder below when answering this question.)

If the IRM Channel 12 range switch is taken to Range 6, what response, if any would be initiated by the Nuclear Instrumentation system?

A. No response B. Rod Block only C. 1/2 Scram only D. Rod Block and 1/2 Scram Answer: B Page: 95 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Answer Explanation:

When on Range 5, the 0 to 40 scale is used. Therefore, IRM Channel 12 is reading approximately 2 counts.

When ranging from Range 5 to Range 6, the scale does not increase. Therefore, 2 on Range 5 equals 2 on Range 6. Range 6 uses the 0-125 scale. A Rod Block occurs when an IRM is downscale (5/125ths of scale).

Distractor 1: Plausible if candidate does not recognize that IRM 14 will be reading downscale.

Distractor 2: Plausible because an APRM downscale with an IRM upscale results in a half scram.

Distractor 3: Plausible if candidate assumes that a 1/2 scram results from being downscale (similar to upscale).

Reference:

QCOP 0700-02 Rev 15, LIC-0702 Rev 8 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Oyster Creek ILT Exam Bank Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

215003.A4.03 (CFR 41.7 / 45.5 to 45.8)

Ability to manually operate and/or monitor in the control room: IRM range switches (RO=3.6 / SRO=3.4)

SR-0702-K21 (Freq: LIC=B)

Given various plant conditions, PREDICT how Intermediate Range Monitor System/plant parameters will respond to manipulation of the following Intermediate Range Monitor System controls:

a. Range switch

b. Mode switch

c. Bypass Joy Stick Page: 96 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 49 ID: QDC.ILT.16455 Points: 1.00 The following conditions exist on Unit 1:

• Startup is in progress.

• Reactor power is STEADY on IRM Range 4.

• SRMs are currently being withdrawn.

While withdrawing the SRMs, the RO notices the SRM Period Meters indicate as shown below.

10 12 P E

15 R I

20 O D

30 S

50 E C

100 O N

D S

-100 Is the SRM response EXPECTED or NOT EXPECTED and why?

(Note: Assume that core flux is peaked at the TOP of the core.)

A. This reading is EXPECTED. The period meter is responding to detector movement and is NOT an accurate indication of period.

B. This reading is NOT EXPECTED. The period meter should be indicating infinity because reactor power is steady.

C. This reading is NOT EXPECTED. The period meter should indicate a positive period because the detectors are being withdrawn from the core.

D. This reading is EXPECTED. The SRMs are currently bypassed.

Answer: A Page: 97 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Answer Explanation:

While withdrawing a SRM from the core, power will appear to drop (negative period) with a top-peaked core.. However, actual power remains steady (actual period is infinity).

Distractor 1: Plausible because reactor power is steady. Incorrect because the flux that the SRMs detect lowers as the SRMs are being withdrawn from the core.

Distractor 2: Plausible because this is the indication expected when inserting the SRMs and also expected if the core was bottom-peaked (flux).

Distractor 3: Plausible if candidate assumes that the SRMs are bypassed when withdrawn similar to other functions (i.e. interlocks) that are bypassed with the current plant conditions.

Reference:

QCOP 0700-01 Rev 12 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.627620)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

SR-0701-K20 (Freq: LIC=B)

Given the following Source Range Monitor System indications/responses and various plant conditions including startup, shutdown, refueling, and scram, EVALUATE the indications/responses and DETERMINE if the indication/ response is expected and normal.

a. SRM/IRM detector position display lights

b. SRM count rate

c. SRM period

d. SRM trip/status lights 215004.A3.01 (CFR 41.7 / 45.7)

Ability to monitor automatic operations of the SOURCE RANGE MONITOR (SRM)

SYSTEM including: Meters and recorders (RO=3.2 / SRO=3.2)

Page: 98 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 50 ID: QDC.ILT.16450 Points: 1.00 Unit 1 was at rated power with SBLC in a normal lineup, when a transient occurred resulting in an ATWS.

Current plant conditions are as follows:

• Reactor power is 43% and steady

• Reactor pressure is 960 psig and steady

• SBLC selector switch was placed in the SYS 1 position Using the 901-5 panel indications shown below, which of the following statements CORRECTLY defines the current operation of the SBLC system?

The 1A SBLC pump is RUNNING...

KEY COLOR means the light is LIT GRAY means the light is NOT LIT A. and INJECTING into the reactor.

B. and pumping to the TEST tank.

C. and pumping to the STORAGE tank.

D. deadheaded with NO flowpath available.

Answer: C Page: 99 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Answer Explanation:

With the SQUIB A light still lit, the squib valve potentially has not fired. This is confirmed with the absence of a FLOW light indication. Therefore, no flowpath to the reactor exists.

Discharge pressure of the pump is at the lifting setpoint of the SBLC pump relief valve.

Therefore, the SBLC pump is pumping to the storage tank via the relief valve.

Distractor 1: Plausible if candidate interprets the open injection valve as a open squib valve and believes indications are normal for injection into the reactor.

Distractor 2: Plausible if candidate assumes that the relief valve discharges to the test tank.

Distractor 3: Plausible if candidate does not recognize that the pump discharge pressure corresponds to the relief valve setpoint, does not remember that there is a relief valve or assumes it is a centrifugal pump. There must be a flowpath available (via a relief valve) because a positive displacement pump will be destroyed if running with no flowpath.

Reference:

QCOP 1100-02 Rev 12, LIC-1100 Rev 17 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

211000.K4.10 (CFR 41.7)

Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: Over pressure protection (RO=2.8 / SRO=3.1)

SR-1100-K22 (Freq: LIC=B)

Given a SBLC operating mode and various plant conditions, PREDICT how SBLC/plant parameters will respond to the following SBLC component or controller failures:

a. Relief Valve fails open

b. Squib Valve fails to open

c. SBLC Storage Tank temperature controller fails on/off Page: 100 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 51 ID: QDC.ILT.16454 Points: 1.00 Regarding the SRM controls at the 901-5 panel:

The operable SRM detectors are normally __(1)__. On a reactor scram, the SRMs will __(2)__.

A. (1) de-selected (2) automatically drive into the reactor core B. (1) de-selected (2) have to be manually driven into the reactor core C. (1) selected (2) automatically drive into the reactor core D. (1) selected (2) have to be manually driven into the reactor core Answer: C Answer Explanation:

All operable withdrawn SRMs should always be left selected. De-selecting a withdrawn operable SRM will prevent automatic insertion of that SRM during a reactor scram Distractor 1: Plausible because the position display button is normally left de-selected.

Distractor 2: Combination of distractor 1 and 3.

Distractor 3: Plausible because the SRMs have to be manually withdrawn from the core.

Reference:

QCOP 0700-01 Rev 12 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.627616)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 101 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-0701-K13 (Freq: LIC=I)

DESCRIBE the following Source Range Monitor System interlocks, including purpose, setpoints, and when/how they are bypassed.

a. Auto insertion

b. Retract permissive 215004.K4.04 (CFR 41.7)

Knowledge of SOURCE RANGE MONITOR (SRM) SYSTEM design feature(s) and/or interlocks which provide for the following: Changing detector position (RO=2.8 /

SRO=2.9)

Page: 102 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 52 ID: QDC.ILT.16433 Points: 1.00 Given the following plant conditions:

• Reactor water level has just reached -59 inches and is continuing to DECREASE

• Drywell pressure is 2.1 psig and STEADY

• Relief Valve 203-3B was manually cycled ONCE to control reactor pressure and its switch is now in the OFF position

• All other relief valve switches are in AUTO

• Reactor pressure is 920 psig and STEADY Assuming NO operator actions, complete the following statement regarding the EARLIEST that ADS will actuate, and the consequences of the actuation.

ADS will actuate in __(1)__ and __(2)__ relief valves will open.

A. (1) 110 seconds (2) FIVE B. (1) 110 seconds (2) FOUR C. (1) 8 1/2 minutes (2) FIVE D. (1) 8 1/2 minutes (2) FOUR Answer: C Answer Explanation:

Without a 2.5 psig drywell pressure signal, ADS initiation will wait 8.5 minutes (-59 inches, 8.5 minutes and >100 psig on discharge of low pressure ECCS pump).

The control switch in OFF will NOT prevent a relief valve from opening on an ADS signal.

Distractor 1: Plausible because if drywell pressure was > 2.5 psig, ADS would initiate in 110 seconds.

Distractor 2: Combination of distractor 1 and 3.

Distractor 3: Plausible because taking the control switch to OFF will defeat the relief function of the valve (vice the ADS function).

Reference:

QCOP 0203-01 Rev 13, QCAN 901(2)-3 G-15 Rev 16 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.03-01 CERT RO 16)

Question History: N/A Page: 103 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Associated objective(s):

SR-0203-K07b (Freq: LIC=B)

LIST the signals which cause an ADS auto initiation including setpoints. DESCRIBE how they are bypassed AND how they are reset.

218000.K5.01 (CFR 41.5 / 45.3)

Knowledge of the operational implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION SYSTEM : ADS logic operation (RO=3.8 /

SRO=3.8)

Page: 104 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 53 ID: QDC.ILT.16513 Points: 1.00 Unit 2 is in the process of a plant startup after a refueling outage. Containment inerting is in progress with oxygen content at 14% and lowering slowly.

In support of this evolution, the "B" Train of the Standby Gas Treatment System (SBGTS) was started MANUALLY and aligned to take a suction on the primary containment.

• An incorrect valve lineup results in reactor water level LOWERING to -10 inches.

Based on this condition, what is the resultant effect on the OXYGEN CONCENTRATION in the primary containment?

The oxygen concentration will...

A. RISE due to the increased flow from SBGTS cycling the Reactor Building-to-Torus vacuum breakers.

B. continue to lower but at a SLOWER rate because MO 2-7504B, U2 TURB BLDG CLG AIR DMPR is OPEN.

C. continue to lower at the SAME RATE because the inerting lineup is NOT affected by the low reactor water level condition.

D. REMAIN CONSTANT because AO 2-1601-63, VENT TO SBGTS is closed.

Answer: D Answer Explanation:

This question tests the candidates knowledge on the impact of a loss of SBGTS on oxygen content.

To lineup drywell inerting, the primary containment isolation valves to the SBGTS system are opened. A low reactor water level signal results in a PCIS Group 2 isolation signal and a SBGTS auto-start signal. The primary containment isolates and the suction valves to the SBGTS system close.

Distractor 1: Plausible if candidate recognizes that the nitrogen makeup isolates on a Group 2 signal but does not recognize that SBGTS system is isolated from the primary containment.

Distractor 2: Plausible because the SBGTS suction to the reactor building will open on an initiation signal. Incorrect because the Turbine Building Suction Damper closes on an auto-initiation signal.

Distractor 3: Plausible if candidate does not recognize that the inerting lineup is affected by a Group 2 - Low RPV water level signal with SBGTS train already running.

Reference:

QCOP 1600-19 Rev 24, QCOP 7500-01 Rev 19, M-76 Rev BJ, M-44 Rev AP, QCAP 0200-10 Rev 41 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Page: 105 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Tier: 2 Group: 1 Question Source: Columbia ILT Exam Bank Question History: Columbia 2003 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Associated objective(s):

261000.K3 06 (CFR 41.7 /45.6)

Knowledge of the effect that a loss or malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following: Primary containment oxygen content: Mark-I&II (RO=3.0 / SRO=3.3)

SR-1602-K20 (Freq: LIC=B)

Given a Primary Containment Atmosphere Control Systems operating mode and various plant conditions, EVALUATE the following Primary Containment Atmosphere Control Systems indications/responses and DETERMINE if the indication/ response is expected and normal.

a. Control Room (1) DW cooler and booster fan status lights (2) Drywell/torus purge fan status lights (3) Nitrogen Purge (Inerting) /Makeup (a) Makeup flow rate (b) Purge flowrate (4) Drywell/torus differential pressure control system (Pumpback)

(a) Drywell/torus differential pressure (b) Joy compressor status lights (5) Primary containment vent valve positions (6) Primary containment hardened vent valve positions (7) Primary Containment Oxygen Analyzer (a) Drywell/torus O2 concentrations (b) Sample point lights

b. Local Page: 106 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 54 ID: QDC.ILT.16457 Points: 1.00 Unit 1 is at 100% power with the CRD system operating in its NORMAL lineup with all HCU accumulators fully charged. The following conditions exist:

• 1A CRD Pump ON

• Charging Water Header pressure 1500 psig and steady

• Drive Water Header pressure 280 psid and steady

• Cooling Water flow 60 gpm and steady

• CRD Flow Controller in MANUAL Complete the statement describing the CRD system flow response as indicated on the 1-340-1 CRD Flow Controller if the MO 1-301-2A, 1A CRD PMP DISCH VLV were throttled open.

INDICATED flow on the CRD Flow Controller will ...

A. increase steadily then stabilize at > 60 gpm.

B. lower steadily then stabilize at < 60 gpm.

C. initially increase then return and stabilize at 60 gpm.

D. initially lower then return and stabilize at 60 gpm.

Answer: A Answer Explanation:

CRD system flow is sensed downstream of the pump discharge valve after which system flow is then directed to either the Charging Water Header or the CRD Flow Control Station (Valves). With the Flow Controller in manual mode, the flow control valve will remain as is and NOT attempt to control at the controller setpoint. No flow is present through the charging water header, therefore, as the CRD pump discharge valve is opened, system flow will rise.

Distractor 1: Opening the pump discharge valve reduces pump discharge pressure, NOT system flow.

Distractor 2: Describes the response with the CRD Flow Controller in Auto.

Distractor 3: Combination of Distractor 1 and Distractor 2.

Reference:

Figure 0302-04 of LIC-0302 Rev 13 Reference provided during examination: No Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Page: 107 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Comments: None Associated objective(s):

201001.A1.03 Ability to predict and/or monitor changes in parameters associated with operating the CONTROL ROD DRIVE HYDRAULIC SYSTEM controls including:

CRD system flow (RO=2.9 / SRO=2.8)

SR-0302-K21 (Freq: LIC=B)

Given a Control Rod Drive Hydraulics operating mode and various plant conditions, PREDICT how key system/ plant parameters will respond to manipulation of the following Control Rod Drive Hydraulics local/remote controls:

a. CRD pump control switch

b. CRD pump discharge valve control switches

c. CRD FCV remote controls (1) Manual/Auto pushbuttons (2) Auto setpoint adjust pushbuttons (3) Manual adjust lever

d. CRD FCV local control switches (1) Air signal selector switch (local/remote)

(2) Signal selector switch (A/B)

(3) Manual regulator knob

e. Stabilizing valve selector switch

f. MO 1(2)-302-8 valve control switch

g. Scram discharge volume vents and drains control pushbuttons

h. Scram discharge volume high level scram bypass switch

i. Scram discharge volume isolation test switch

j. HCU alarm panel alarm light pushbuttons and panel test switch Page: 108 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 55 ID: QDC.ILT.16460 Points: 1.00 Unit 2 is operating at rated power when the following occurs:

• 912-1 D-1, RX BUILDING COOLING WATER LOW PRESSURE, alarms.

• 902-4 A-17, DRYWELL FLOOR DRAIN SUMP HIGH LEVEL, alarms.

• Drywell pressure increases from 1.24 psig to 1.40 psig.

As directed by the Unit Supervisor, the Admin NSO closes the following valves:

• MO 2-3702, U2 RBCCW SPLY

• MO 2-3703, U2 RBCCW OUTBD RTN VLV

• MO 2-3706, U2 RBCCW INBD RTN VLV With NO further operator actions, which of the following statements is correct?

A. Recirculation Pump Seal temperatures will rise.

B. Fuel Pool temperature will rise.

C. RWCU Pump bearing temperatures will rise.

D. Pumpback (Joy) air compressor temperature will rise.

Answer: A Answer Explanation:

Answer: Isolating RBCCW to the Drywell by closure of the three valves will result in a loss of cooling flow to the Recirc Pump seals, (QCOA 3700-01).

Distractor 1: Loss of RBCCW to the Drywell Coolers result in any damage to the fan or components. QCOA 3700-06, RBCCW Line Break Inside Containment directs the operator to continue running the Drywell Coolers.

Distractor 2: Closure of the 3 RBCCW valves will not isolate cooling flow to the Fuel Pool Cooling heat exchangers as these are outside of containment.

Distractor 3: Closure of the 3 RBCCW valves will not isolate cooling flow to the Joy aftercooler as this are outside of containment.

Reference:

QCOA 3700-01 Rev 7, LNF-3700 Rev 00 Reference provided during examination: No Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: Bank Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 109 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-3700-K22 (Freq: LIC=B)

Given a RBCCW operating mode and various plant conditions, PREDICT how RBCCW/plant parameters will respond to the following RBCCW failures:

a. Leak into RBCCW

b. Leak from RBCCW (inside/outside the drywell)

c. LCV fails open or shut

d. Loss of service water flow

e. Pump trip 202001.K6.02 Knowledge of the effect that a loss or malfunction of the following will have on the RECIRCULATION SYSTEM : (CFR 41.7 / 45.7)

Component cooling water systems (RO=3.1 / SRO=3.2)

Page: 110 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 56 ID: QDC.ILT.16464 Points: 1.00 The Reactor Water Cleanup System (RWCU) is in normal operation at full power. Panel 901-4 indications are provided below.

(Note: The 1-1290-28, RWCU LOOP TEMP SELECT switch has been partially obscured in the picture.)

Based on the indications provided, which point is the 1-1290-28, RWCU LOOP TEMP SELECT switch measuring?

RWCU LOOP TEMP SELECT 1-1290-28 POINT RWCU WATER TEMPERATURES 1 WTR TO RWCU SYSTEM 2 REGEN HX OUTLET 3 NON-REGEN HX OUTLET 4 RETURN TO FEEDWATER LINE 5 SPARE 6 SPARE A. 1 B. 2 C. 3 D. 4 Answer: A Page: 111 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Answer Explanation:

Answer: The provided picture is with the switch in position 1. Meter 1-1290-21 is reading approximately 530ºF, which is close to saturation temperature of the reactor water. The RWCU Regen HX cools the reactor water from 540-550ºF. Therefore, meter 1-1290-21 is indicating WTR TO RWCU SYSTEM, Point 1.

Distractor 1: Normal Pt 2 value is 210-220 ºF.

Distractor 2: Normal Pt 3 value is90-110 ºF.

Distractor 3: Normal Pt 4 value is 450ºF.

Reference:

LN-1200, Rev. 1 Reference provided during examination: N/A Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.699575)

Question History: Modified version of it on 2009 ILT NRC Exam 10 CFR Part 55 Content: 41.7 Comments: None Associated objective(s):

204000.K5.08 Knowledge of the operational implications of the following concepts as they apply to REACTOR WATER CLEANUP SYSTEM : (CFR 41.5 / 45.3)

Temperature measurement (RO=2.6 / SRO=2.6)

SR-1200-K20 (Freq: LIC=B) Given a Reactor Water Cleanup System operating mode and various plant conditions, EVALUATE the following Reactor Water Cleanup System indications/responses and DETERMINE if the indication/ response is expected and normal:

a. Control Room (1) RWCU recirc pump discharge pressure (2) Filter-demin flow (3) System temperatures (4) Reject flowrate (5) RWCU conductivities and dissolved O2 (6) RWCU recirc pump status (7) MOV and AOV positions Page: 112 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 57 ID: QDC.ILT.16466 Points: 1.00 A control rod has been withdrawn one notch and has latched into the desired notch position.

Which statement describes how the control rod position is detected for the Full Core Display on the 901-5 panel?

A. The even reed switch at this position is closed.

B. The even reed switch at this position is open.

C. The odd reed switch at this position is closed.

D. The odd reed switch at this position is open.

Answer: A Answer Explanation:

Answer: Control rods latch at even reed positions. Reed switches close when the control rod is at the respective position.

Distractor 1: Reed switches close to indicate position.

Distractor 2: Control rods only latch at even positions, the odd reed switch should be open and not indicating.

Distractor 3: Combination of Distractors 1 and 2.

Reference:

UFSAR Section 7.7 Revision 11, LIC-0280 (Reactor Manual Control and Rod Position Information Systems) Rev.12.

Reference provided during examination: No Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: Bank Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification:

Comments: None Page: 113 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-0280-K14 (Freq: LIC=B)

STATE the physical location and function of the following principal Reactor Manual Control System (RMCS)/ Rod Position Information System (RPIS) components:

a. Rod Position Information System (1) RPIS power supply and fuses (901(2)-27)

(2) RPIS probe buffer cards/RPIS input cable and connector (901(2)-27)

(3) RPIS reed switches

b. Reactor Manual Control System (1) Rod select matrix (2) Automatic sequence timer (3) Auxiliary timer (4) Rod drift alarm circuit (5) Refueling interlocks (6) Rod block jumper points (CRD accummulator low pressure / SRM/IRM FULL IN interlock) (901(2)-28)

c. Standby Sequence Timer 214000.K4.01 Knowledge of ROD POSITION INFORMATION SYSTEM design feature(s) and/or interlocks which provide for the following: (CFR 41.7)

Reed switch locations (RO=3.0 / SRO=3.1)

Page: 114 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 58 ID: QDC.ILT.16470 Points: 1.00 (Refer to the portion of M-37, Diagram of RHR Service Water Piping, provided.)

RHR Loop A is operating in Torus Cooling mode on Unit 1.

• Torus water temperature is LOWERING slowly.

If RV 1-1001-165A, RHR HX Service Water Relief Valve, were to OPEN, what effect, if any, would this have on the Torus Cooling system performance and why?

(Assume NO Operator Actions are taken)

The Torus water temperature COOLDOWN RATE will...

A. RISE because of INCREASED RHR Service Water flow through the heat exchanger.

B. LOWER because of REDUCED RHR Service Water flow through the heat exchanger.

C. REMAIN CONSTANT because RHR system flow is unchanged.

D. REMAIN CONSTANT because the relief valve is on the outlet of the heat exchanger.

Answer: B Answer Explanation:

Diverting RHR Service Water Flow to the RHR Heat Exchanger will lower the heat removal capability if no adjustments are made. Therefore, Torus water temperature cooldown rate will lower.

Distractor 1: RHR Service Water flow is decreased with the RHR Service Water Reversing Valve lineup valve lineup as shown, resulting in a smaller Torus temperature cooldown rate.

Distractor 2: Heat Exchanger heat removal capability is diminished resulting in a reduced Torus temperature cooldown rate.

Distractor 3: The relief valve is at the inlet which reduces RHR Service Water flow to the Heat Exchanger.

Reference:

P&ID M-37 Rev BC Reference provided during examination: Yes Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: Bank Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

Page: 115 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

SRO Justification: N/A Comments: None Associated objective(s):

219000.K3.01 Knowledge of the effect that a loss or malfunction of the RHR/LPCI:

TORUS/SUPPRESSION POOL COOLING MODE will have on following: (CFR 41.7 /

45.4)

Suppression pool temperature control (RO=3.9 / SRO=4.1)

SR-1000-K14 (Freq: LIC=B)

STATE the physical location and function of the following principle RHR/RHRSW system components:

a. RHR pumps

b. RHR MOVs

c. RHR CCST/Torus manual suction valves

d. Manual valves for RHR crosstie flowpaths to condenser, condensate pump or Radwaste

e. RHRSW pumps

f. RHRSW MOVs

g. RHRSW unit cross-tie manual valves

h. RHR and RHRSW vault doors

i. RHRSW vault sump pump

j. RHR room cooler and control panel

k. RHRSW cubicle cooler

l. 901(2)-32 panel:including test jack, keylock switches, Group 2 relay, fuses and terminals for jumpering S17,18,19.

m. 901(2)-40 panel:including SDC low pressure interlock relay, Group 2 relay, and terminals for 1001-50 valve jumpers.

n. 901(2)-33 panel:including leads lifted to keep RHRSW-5 valve open

o. RHR check valve AO-1001-68A/B

p. Integrated Leak Rate Test (ILRT) connection and spool piece

q. RHR minimum flow valve manual stop valve

r. RHR heat exchanger manual outlet valve (17A/B)

s. Connections for RHR heat exchanger level indicating tygon tubes (RHR side/RHRSW side)

t. RHR heat exchanger vents/drains (RHR side/RHRSW side)

u. RHR heat exchanger RHRSW side relief valve (1(2)-1001-165A/B)

v. Safe shutdown hose and 1(2)-1099-166 / 1(2)-4199-291 valves

w. Thermally induced pressurization relief valve between 47 and 50

x. 901(2)-32/33 panel terminals for bypassing drywell and torus spray interlock signals (901(2)-32/33)

y. MO-1(2)-1001-47 250vdc disconnect switch (1(2)-1001-DS)

Page: 116 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 59 ID: QDC.ILT.16473 Points: 1.00 The following conditions exist on Unit 1 during a LOCA:

• RPV pressure is 800 psig and lowering.

• Drywell pressure is 8.0 psig and lowering.

• Torus sprays are operating on B Loop of RHR

• Drywell sprays are operating on A Loop of RHR.

• Both loops of RHR are cross-tied.

Three minutes later, Transformer 12 trips and the NSOs reports the following:

• The Unit 1 Emergency Diesel Generator did NOT start.

If all other automatic actions occurred as designed, select the correct statement describing RHR system status.

A. The 1C and 1D RHR pumps are TRIPPED.

The 1A and 1B RHR pumps are SPRAYING the drywell and torus.

B. The 1C and 1D RHR pumps are TRIPPED.

The 1A and 1B RHR pumps are SPRAYING the drywell ONLY.

C. The 1B and 1D RHR pumps are TRIPPED.

The 1A and 1C RHR pumps are SPRAYING the drywell and torus.

D. The 1B and 1D RHR pumps are TRIPPED.

The 1A and 1C RHR pumps are SPRAYING the drywell ONLY.

Answer: A Answer Explanation:

With the loss of offsite power resulting from the trip of Transformer 12, the Emergency Diesel Generators will auto start and supply power to the ECCS Busses. With the failure of the Unit 1 EDG, Bus 14-1 is de-energized and therefore 1C and 1D RHR pumps are tripped. The RHR torus spray valves are MO valves, therefore the remain open when power is lost.

Distractor 1: Plausible if candidate assumes that Torus sprays secure on the B loop when power is lost.

Distractor 2: Plausible if candidate assumes that the Division 2 RHR pumps are the 1B and 1D (notation similar to other systems).

Distractor 3: Combination of distractor 1 and 2.

Reference:

QOM 1-6500-T06 Rev 7 Reference provided during examination: No Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: New Page: 117 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None.

Associated objective(s):

230000.K2.02 (CFR 41.7)

Knowledge of electrical power supplies to the following: Pumps (RO=2.8 / SRO=2.9)

SRN-6500-K24 (Freq: LIC=B NF=B) Given a 4KV / 480 VAC Distribution Systems operating mode and various plant conditions, PREDICT how each supported system will be impacted by the following 4KV / 480 VAC Distribution Systems failures:

a. Loss of T11 and/or T12 (T21/22)

b. Loss of a 4KV bus

c. Loss of a 480 VAC bus

d. Loss of a 480 VAC MCC

e. T12/22 regulator fails HIGH or LOW Page: 118 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 60 ID: QDC.ILT.16476 Points: 1.00 Fuel is being offloaded to the Fuel Pool when a complete loss of RBCCW to the Fuel Pool Cooling (FPC) system occurs.

Which of the following methods, (without the use of hoses, spool pieces, or other temporary connections), could be used to reject decay heat and maintain Fuel Pool temperature in band?

A. Fuel Pool Cooling Assist using RHR Service Water.

B. Feed and Bleed using Condensate Transfer.

C. Feed and Bleed using the Fire System.

D. Feed and Bleed using Clean Demin.

Answer: B Answer Explanation:

Answer: Condensate Transfer is the normal means of adding water to the Fuel Pool system. Opening the 1(2)-1901-112 valve will add water directly to the Skimmer Surge Tank.

Distractor 1: An acceptable method of water addition, but requires a hose connection from the tap to the Fuel Pool.

Distractor 2: Employs the use of the RHR system, but requires placement of a spool piece to connect the Fuel Pool to the RHR System.

Distractor 3: Same as distractor 1.

Reference:

QCOA 1900-02, Rev.6 Reference provided during examination: No Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(2) to 41(b)(9)

SRO Justification:

Comments: None.

Page: 119 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

233000.K1.16 Knowledge of the physical connections and/or cause- effect relationships between FUEL POOL COOLING AND CLEAN-UP and the following: (CFR 41.2 to 41.9 /

45.7 to 45.8)

Emergency cooling water systems: Plant-Specific (RO=2.7 / SRO=2.8)

SR-1000-K19 (Freq: LIC=B)

LIST the plant systems which support the RHR system and DESCRIBE the nature of support. (Includes power supplies)

Page: 120 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 61 ID: QDC.ILT.16479 Points: 1.00 Unit 1 is in Mode 1.

Which of the following instruments, if determined to be INOPERABLE, will affect instrumentation controlled by TS LCO 3.3.1.1, Reactor Protection System (RPS)?

A. LSH 1-0263-25-3A, Reactor Vessel High Level.

B. LSH 1-3541-37C, Moisture Separator High Level.

C. PIS 1-0263-22A, Recirculation Pump Trip - High Reactor Pressure.

D. PS 1-5641-122, Turbine Control Valve Fast Closure Trip Oil Pressure-Low.

Answer: D Answer Explanation:

Answer: Low Oil Pressure RPS trip, (> 475 psig) is required by TS Table 3.3.1.1-1, Function 9.

Distractor 1: Reactor high water level trip (50.34 in.) is required by Tech Specs 3.3.2.2, Feedwater and Main Turbine Instrumentation, NOT, RPS.

Distractor 2: Moisture Separator high level trip of the Main Turbine is NOT required by Tech Specs.

Distractor 3: Recirc Pump Trip on high Reactor pressure (1250 psig) is required by TS LCO 3.3.4.1 (ATWS - RPT).

Reference:

QCAN 901-5 A-13, Channel A/B Turb-Gen Load Mismatch Low EHC Press, Rev.11 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 121 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-0500-K29 (Freq: LIC=I)

Given Reactor Protection System key parameter indications and various plant conditions, DETERMINE, from memory, if the Reactor Protection System Tech Spec LCOs have been met.

241000.2.2.42 Reactor/Turbine Pressure Regulating System:

Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (RO=3.9 / SRO=4.6) specifications.

Page: 122 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 62 ID: QDC.ILT.16484 Points: 1.00 Unit 1 is at 90% power. A special test on the DEHC system is in progress.

Current plant conditions are as follows:

• Actual Turbine Load = 85%

• Reactor pressure = 973 psig

• Turbine Load Set = 100%

• Dome Pressure Set = 915 psig

• Throttle Pressure Set = 906 psig

• DEHC mode = Dome Pressure Control

• All other DEHC controls are NORMAL Which of the following operator actions will result in a BYPASS valve OPENING?

(Note: Consider each action separately.)

A. LOWER Max Combined Flow Limiter to 80%.

B. LOWER Turbine Load Set to 80%.

C. LOWER Dome Pressure Set to 910 psig.

D. LOWER Throttle Pressure Set to 900 psig.

Answer: B Answer Explanation:

Turbine Load Set limits the turbine control valve position to correspond to a load limit on the turbine. Therefore, lowering Turbine Load Set to 80% will close the turbine control valves to lower turbine load. Since reactor power has not changed, a bypass valve will open to pass the "extra" 5% of steam to the main condenser.

Distractor 1: Lowering MCFL will close the turbine control valves, but will also prevent the bypass valves from opening (MCFL limits the combined total of steam flow through the bypass valves and control valves). Reactor pressure will rise until the scram setpoint is reached.

Distractor 2: Lowering Dome Pressure set to 910 psig will result in an opening of the Turbine Control valves (since Turbine load set is at 100%). This will result in an increase in turbine loading and a reduction in steam dome pressure. Bypass valves will not open.

Distractor 3: Lowering Throttle Pressure set would have the same effect as lowering Dome Pressure set if DEHC was in Throttle Pressure Control mode of operation.

Reference:

LIC-5652a Rev 5 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: Modified from Quad Cities ILT Bank (QDC.ILT.109083)

Page: 123 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Question History: N/A 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None.

Page: 124 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

245000.A4.07 (CFR 41.7 / 45.5 to 45.8)

Ability to manually operate and/or monitor in the control room: Turbine valve position (RO=2.9 / SRO=2.9)

SR-5652a-K21 (Freq: LIC=B) Given a Main Turbine Control - EHC Logic System operating mode and various plant conditions, PREDICT how Main Turbine/EHC systems and plant parameters will respond to manipulation of the following Main Turbine Control -

EHC Logic System local/remote controls:

a. Status (1) Turbine Reset (2) Pressure Reset (3) TB Diag Reset (4) PP Diag Reset

b. Control (1) Valve Limiters (a) Load Limit (b) Max Combined Flow Limit (2) Pressure Control (a) Pressure Control Mode - Throttle/Dome (b) Pressure Set (3) Pre-Warming (a) MSV #2 adjustment (b) Chest Warming (c) Shell Warming (4) Speed-Load (a) Speed Control

1) Acceleration

2) Speed Cmd (b) Load Set (5) Rx Cooldown (a) OFF/ON (b) Setpoint/Rate (6) BPV Jack

c. Aux (1) Trip Status (a) 1st Hits (b) Disable/Enable Hi Vibe Trip (2) Mark VI Panel (a) Comm Status (b) Power Supplies (3) Transmitter Reset

d. Tests (1) PCU (2) MSV-CV (3) BPV (4) CIV (5) 2/3 Trip System (6) Offline Vlv Test (7) Offline O/S (8) Online O/S

e. Trubine trip pushbuttons Page: 125 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 63 ID: QDC.ILT.16490 Points: 1.00 (A reference is provided for this question.)

A FAS alarm is received in the Control Room.

• FAS Terminal Display indicates FAS Device "43-30" in alarm.

An EO is dispatched and reports from the local protectowire panel:

• Zone 2 alarm and the footage meter indicates 210.

Where is the potential fire?

A. 1A RHR Room B. 1B RHR Room C. 2A RHR Room D. 2B RHR Room Answer: A Answer Explanation:

Answer: Attachment A shows that FAS device 43-30 is associated with Unit 1 RB 554' North (local panel 2201-81). Zone 2 (footage 0-226) on Attachment V associates alarm with the 1A RHR Room Cable Pan and Risers.

Distractor 1, 2 and 3 are incorrect: Plausible because the student can use the footage reading on the incorrect Zone.

Distractor 2 is incorrect: Combination of Distractor 1 and 3 Distractor 3 is incorrect: Plausible because the student can use the footage reading on the incorrect Unit.

Reference:

QCOA 4100-11 Rev 22 Reference provided during examination: QCOA 4100-11 Rev 22 (all)

Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: Modified Question History: 2009 NRC Exam (QDC.ILT.15531) 10 CFR Part 55 Content: 41.7 Comments: None Page: 126 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-4100-K06 (Freq: LIC=I)

Given a Fire Protection Systems annunciator tile inscription OR local panel alarm/trouble light label, DESCRIBE the condition causing the alarm and any automatic actions which occur when the alarm actuates. EXPLAIN the consequences of the condition if not corrected.

286000.A3.04 Ability to monitor automatic operations of the FIRE PROTECTION SYSTEM including: (CFR 41.7 / 45.7)

System initiation (RO=3.2 / SRO=3.3)

Page: 127 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 64 ID: QDC.ILT.16493 Points: 1.00 Unit 1 was operating at rated power with Control Room Ventilation in its NORMAL lineup when a LOCA occurred. Current conditions are:

• Drywell pressure is 4.0 psig and slowly rising

• RPV water level is zero (0) inches and steady

• Transformer 12 is supplying Busses 13 and 14

• All automatic actions occurred.

Select the statement describing the status of Control Room ventilation and what actions are required.

A. The "A" Train has tripped. Start the "B" Train.

B. The "A" Train is operating in Recirculation mode and the AFU must be started within 40 minutes.

C. The "B" Train is operating in Recirculation mode and the AFU must be started within 40 minutes.

D. The "A" Train is operating with a normal damper lineup. Place the CONTROL ROOM HVAC ISOL switch to ISOLATE.

Answer: B Answer Explanation:

Control Room ventilation isolates and operates in the Recirculation mode on a LOCA signal, ie. high Drywell pressure or low RPV water level. Since there was no loss of offsite power, the "A" Train would continue to operate in the Recirc mode. The AFU must be started within 40 minutes of a LOCA per QCOP 5750-09.

Distractor 1: The "A" Train does not trip because offsite power is not lost. Plausible because it isolates which one may incorrectly assume the system trips.

Distractor 2: Same as Distractor 1, except that "B" Train would autostart in this case.

Distractor 3: Plausible because "A" Train does remain in operation, however it isolates automatically. Manual isolation is not necessary.

Reference:

QCOP 5750-09, Rev. 44 Reference provided during examination: No Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Page: 128 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

288000.A2.01 (CFR 41.5 / 45.6)

Ability to (a) predict the impacts of the following on the PLANT VENTILATION SYSTEMS

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

High drywell pressure: Plant-Specific (RO=3.3 / SRO=3.4)

SR-5752-K26 (Freq: LIC=B)

EVALUATE given key Control Room Ventilation System parameter indications and/or responses depicting a system specific abnormality/ failure and DETERMINE a course of action to correct or mitigate the following abnormal condition(s):

a. Complete loss of control room ventilation

b. Toxic gas/release from nearby chemical plant Page: 129 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 65 ID: QDC.ILT.16496 Points: 1.00 The following plant conditions exist on Unit 1:

Rx Bldg to Atmos DP = - 0.2 in. H2O Torus Pressure = 0.01 psig Drywell Pressure = 1.4 psig Complete the following statements:

If 1-1601-32A, DRYWELL TO TORUS VACUUM BKR is opened at panel 2251-24, Drywell to Torus d/p will __(1)__.

If AO 1-1601-20A, RX BLDG TO TORUS VACUUM BKR is opened at panel 901-3, Torus to Reactor Building d/p will __(2)__.

(Note: Consider the effect of each action separately.)

A. (1) lower (2) remain constant B. (1) lower (2) lower C. (1) remain constant (2) remain constant D. (1) remain constant (2) lower Answer: A Answer Explanation:

Answer: If a Drywell to Torus vacuum breaker is opened, a direct path is provided from the torus to the drywell. If a Rx Bldg to Torus vacuum breaker is open, a check valve will maintain primary containment integrity up to 0.5 psid between the reactor building and torus. Under normal conditions the reactor building to torus differential pressure varies between .1 to .15 psid.

Distractor 1: Check valve, 1-1601-31A, maintains primary containment integrity.

Distractor 2: Drywell will relieve to the Torus.

Distractor 3: See above

Reference:

QCOS 1600-01, Rev. 17, QCOS 1600-14, Rev.21, M-34 Sh.1 Reference provided during examination: No Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 Question Source: Bank (QDC.ILT.627862)

Question History: N/A Page: 130 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 10 CFR Part 55 Content: 41(b)(5)

SRO Justification: N/A Comments: None Associated objective(s):

290001.A1.01 Ability to predict and/or monitor changes in parameters associated with operating the SECONDARY CONTAINMENT controls including: (CFR 41.5 / 45.5)

System lineups (RO=3.1 / SRO=3.1)

SR-1601-K21 (Freq: LIC=B)

Given various plant conditions, PREDICT how key system/ plant parameters will respond to manipulation of the following Containment Systems controls:

a. Reactor Building to torus vacuum breaker control switches

b. Area steam leak detection monitor control switches.

Page: 131 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 66 ID: QDC.ILT.16472 Points: 1.00 Which of the following temporary directives can NOT be issued and controlled by use of a STANDING ORDER?

A. Describe the preferred way of operating a system to equalize equipment run times based on engineering data.

B. Establish a safety requirement to use a spotter when operating a chainwheel operated component.

C. Require an increased monitoring surveillance to be utilized when remote indication of a radiation monitor is lost.

D. Change the order of performance steps to a test procedure for installation of a new digital control system.

Answer: D Answer Explanation:

Standing Orders shall NOT be used as a substitute for procedures or the daily site schedule. Temporary changes to procedures (including test procedures) are controlled by AD-AA-101, Processing of Procedures and T&RMs which require specific approvals different than Standing Orders.

Distractor 1, 2 and 3: All acceptable directives in accordance with OP-AA-102-104 and all are current or previous standing orders.

Reference:

OP-AA-102-104 Rev 1, AD-AA-101 Rev 22 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(10)

SRO Justification: N/A Comments: None Page: 132 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

2.1.15 Knowledge of administrative requirements for temporary management directives, such as standing orders, night orders, operations memos, etc. (RO=2.7 / SRO=3.4)

SRNLF-PR-K05 (Freq: LIC=I NF=I)

Given corporate and station administrative procedures and an operating situation related to procedure use, EVAUATE the conditions and DETERMINE if a Temporary Procedure Change, Equipment Status Tag, Safety Screening, or Procedure Revision is needed.

Page: 133 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 67 ID: QDC.ILT.16474 Points: 1.00 Unit 2 is in a refuel outage with the reactor vessel head removed and fuel in the vessel.

Which one of the following actions is a CORE ALTERATION per Technical Specifications?

(Consider each action separately.)

A. Movement of irradiated fuel within the fuel pool.

B. Removing a fuel bundle on the peripheral of the core.

C. Installing a control rod blade into an empty cell.

D. Driving a Source Range Monitor detector fully into the core.

Answer: B Answer Explanation:

Core alterations shall be the movement of any fuel, sources, or reactivity control components, within the reactor vessel with the vessel head removed and fuel in the vessel.

Moving a peripheral fuel bundle within the core is a core alteration.

Distractor 1: Plausible because moving irradiated fuel within the vessel is a core alteration.

Distractor 2: Plausible because inserting a control rod into a cell with fuel is an alteration.

Distractor 3: Plausible because SRMs are made from fissionable material.

Reference:

TS Definitions Section 1.1 Amendment No. 233/229 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 Question Source: Susquehanna ILT Exam Bank Question History: Susquehanna 2005 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(2,10)

SRO Justification: N/A Comments: None Page: 134 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

2.1.41 Knowledge of the refueling process. (RO=2.8 / SRO=3.7)

Given the Improved Technical Specifications (ITS) and the associated Bases, the trainee shall:

IDENTIFY THE TECHNICAL SPECIFICATION DEFINITIONS LISTED IN SECTION 1.1 OF THE ITS.

Page: 135 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 68 ID: QDC.ILT.16475 Points: 1.00 Given:

• The plant is in operating in Mode 1.

• A maintenance visual inspection requires momentarily placing an ECCS component, with auto-start capability, in PULL-TO-LOCK.

Complete the following statement.

Per QAP 0300-02, Conduct of Shift Operations, the earliest this ECCS component is OPERABLE is when...

A. it is manually started.

B. its auto-start function is tested.

C. its monthly surveillance is performed.

D. its control switch is returned to normal.

Answer: D Answer Explanation:

Returning a PULL-TO-LOCK control switch to normal immediately makes the component operable.

Distractor 1, 2 and 3: All distractors are plausible because each is a correct answer based on a different degree of maintenance to the system.

Reference:

QAP 0300-02 Rev 69 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A Question Source: Quad Cities Exam Bank Question History: N/A 10 CFR Part 55 Content: 41(b)(10)

SRO Justification: N/A Comments: None Page: 136 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SRNLF-00-K10 (Freq: LIC=I NF=I)

STATE the administrative requirements which, by their nature, must be memorized (for example - color coding of control switches).

2.2.36 Ability to analyze the affect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (RO=3.1 / SRO=4.2)

Page: 137 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 69 ID: QDC.ILT.16478 Points: 1.00 What impact does losing 125 VDC control power have on a CLOSED 4KV safety-related load breaker?

A. Prevents remote and local breaker operation. Breaker protective relaying is REMOVED.

B. Prevents remote breaker operation and breaker protective relaying is REMOVED. Local manual trip capability remains AVAILABLE.

C. Prevents remote breaker operation. Breaker protective relaying and local manual trip capability remain AVAILABLE.

D. Prevents remote and local breaker operation. Breaker protective relaying remains AVAILABLE.

Answer: B Answer Explanation:

Control power is necessary to electrically operate the breaker remotely. Since the breaker is closed, the trip coils are charged and trip capability remains using the local pushbutton (mechanical trip). The protective relaying requires control power to function.

Distractor 1: Plausible if the candidate does not realize that the opening springs are charged when the breaker is closed.

Distractor 2: Plausible if the candidate assumes that the protective relaying is from AC power within the cubicle.

Distractor 3: Combination of distractor 1 and 2.

Reference:

LN-6500 Rev 17 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A Question Source: Quad Cities ILT Exam Bank (QDC.ILT.15564)

Question History: Quad Cities 2009 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(7)

SRO Justification: N/A Comments: None Page: 138 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

2.2.37 Ability to determine operability and/or availability of safety related equipment.

(RO=3.6 / SRO=4.6)

SRN-6500-K23 (Freq: LIC=B NF=B) Given a 4KV / 480 VAC Distribution Systems operating mode and various plant conditions, PREDICT how the 4KV / 480 VAC Distribution Systems will be impacted by the following support system failures:

a. Loss of 125 VDC control power

b. Turbine/generator trip

c. Loss of 345KV bus Page: 139 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 70 ID: QDC.ILT.16482 Points: 1.00 Refer to the weekly surveillance log shown below:

Where does the operator ACCESS the Radwaste Effluent Radiation Monitor reading to determine if it is within weekly surveillance requirements?

A. At a recorder on the 901-4 panel B. At a recorder on the 901-2 panel C. At a SPING terminal on the 912-4 panel D. At a radiation monitor on the 901-10 panel Answer: C Answer Explanation:

The Radwaste Effluent Radiation Monitor is an input into the Eberline SPING terminal on the 912-4 panel.

To determine reading, depress DATA, 10-01, ENTER.

Distractor 1: Plausible because Radwaste pumps are operated from the 901-4 panel.

Distractor 2: Plausible because other process radiation recorders are located on the 901-2 panel.

Distractor 3: Plausible because other process radiation monitors are located on the 901-10 panel.

Reference:

QOS 0005-01 Rev 135 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(11)

SRO Justification: N/A Page: 140 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Comments: None Associated objective(s):

SR-1701-K05 (Freq: LIC=I)

Given the following Process Radiation Monitoring System key parameters, STATE the physical location of the lights, meters, and recorders and DESCRIBE the radiation detector locations in the system flowpath:

a. Main Steam Line Radiation Monitors (1) Radiation levels (2) NUMAC chassis trip/status indications

b. SJAE Monitors (Off-Gas Log Scale)

(1) Radiation levels (2) NUMAC chassis trip/status indications (3) Interval timer status

c. Off-Gas Flux Tilt (Off-Gas Linear Scale) radiation levels

d. Off-Gas Filter Building Process radiation levels and alarm lights

e. Process Liquid Radiation Monitor System (1) Service water radiation levels (2) RBCCW radiation levels and alarm lights (3) Radwaste effluent radiation levels

f. Reactor Building Vent / Fuel Pool Radiation Monitors (1) Indicator/Trip unit radiation levels and Hi/Lo alarm lights (2) Power supply voltage and power on light 2.3.05 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. (RO=2.9 /

SRO=2.9)

Page: 141 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 71 ID: QDC.ILT.16481 Points: 1.00 Complete the following statement regarding Hydrogen Water Chemistry (HWC) system operation:

Allowing the U-1 HWC system to remain in standby (main steam flow < 20% AND the system NOT tripped)...

A. IS permitted to allow automatic start of hydrogen injection.

B. is NOT permitted due to the potential to cause a CRUD burst within the primary system.

C. is NOT permitted due to the potential increase in worker RADIATION exposure levels.

D. is NOT permitted due to the potential of creating an EXPLOSIVE mixture of hydrogen and oxygen within the Offgas system.

Answer: C Answer Explanation:

Hydrogen injection will stop and the system will go into standby whenever main steam flow is < 20% and system is not tripped (this trip is only active after steam flow has been above 20% for one hour or above 25%). Operation in standby mode is not permitted due to the potential increase in worker exposure levels by automatic restart of hydrogen injection if reactor steam flow reaches 20% or greater.

Distractor 1: Plausible because procedure directs operating the HWC system to the greatest extent possible when above 20% reactor power.

Distractor 2: Plausible because hydrogen water chemistry will change the primary system chemistry content.

Distractor 3: Plausible because hydrogen levels in the Offgas system will increase upstream of the recombiner.

Reference:

QCOP 2700-01 Rev 37 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A Question Source: New Question History: N/A 10 CFR Part 55 Content: 41(b)(12)

SRO Justification: N/A Comments: None Page: 142 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-2700-K28 (Freq: LIC=I)

EXPLAIN the reasons for given Hydrogen Water Chemistry System operating limits and precautions.

a. Operation of HWCS is standby is NOT allowed

b. Reactor startup not allowed to continue above 284 degrees until O2 is below 300 ppb

c. Hydrogen piping pressure limit and limit on rate of change

d. Precautions for working with cyrogenic gas

e. Precautions for hydrogen gas

f. Precautions for oxygen enriched atmosphere

g. Precautions for threaded connections onto hydrogen piping 2.3.12 Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. (RO=3.2 / SRO=3.7)

Page: 143 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 72 ID: QDC.ILT.16483 Points: 1.00 While executing a Quad Cities Emergency Operating Procedure (QGA), what must be done if ANOTHER entry condition occurs OR the initial entry condition re-occurs?

A. IMMEDIATELY re-enter the emergency operating procedure.

B. Continue from where you are presently in the procedure, re-entry is NOT required.

C. Wait until you can exit the current leg, THEN re-enter the procedure with the current conditions.

D. Concurrently enter the applicable Severe Accident Management Guideline (SAMG) procedure.

Answer: A Answer Explanation:

If a procedure is being executed and another entry condition occurs, the procedure is re-entered and executed. Action steps for which the intent of the step has already been met and repetition of the step will serve no purpose and do not need to be re-performed.

Distractor 1: .Plausible if candidate assumes that all actions applicable for each entry condition have been completed and re-entry is not necessary.

Distractor 2: Plausible if candidate assumes that QGAs are not executed concurrently Distractor 3: Plausible if candidate assumes that a second entry condition signifies a more severe accident situation.

Reference:

QCAP 0200-10 Rev 41 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A Question Source: Dresden ILT NRC Exam Bank Question History: Dresden 2008 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(10)

SRO Justification: N/A Comments: None Page: 144 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

2.4.14 Knowledge of general guidelines for EOP usage. (RO=3.8 / SRO=4.5)

SR-0001-K05 (Freq: LIC=I)

DESCRIBE the procedure use guidelines contained in QCAP 0200-10

a. QGA procedure entry

b. QGA execution / concurrent use (QGA, QCOA, QCOP, and QCGP)

c. QGA procedure exit Page: 145 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 73 ID: QDC.ILT.16486 Points: 1.00 Which of the following actions can be performed FROM MEMORY for a QCARP during an Appendix R fire?

A. Place the AUTO BLOWDOWN INHIBIT switch to INHIBIT.

B. Close ALL inboard and outboard MSIVs.

C. Verify the REACTOR MODE switch is in RUN.

D. Place all four RFP control switches in PULL-TO-LOCK.

Answer: C Answer Explanation:

OP-AA-101-111, Roles and Responsibilities of On-Shift Personnel, states that during transient conditions, the RO may perform immediate operator actions of abnormal procedures from memory, while verbalizing actions being taken to the Unit Supervisor.

Although this is specifically testing the Appendix R abnormal procedures, the concept of "immediate operator actions" is generic for all abnormal procedures. Focusing on a QCARP abnormal procedure raises the difficulty level while remaining independent of specific system abnormal procedures.

Verifying the Reactor Mode switch is in RUN is an immediate operator action in the QCARPs.

Distractor 1, 2 and 3: Plausible because all actions are subsequent operator actions.

Reference:

OP-AA-101-111 Rev 5, QCARP 0010-01 Rev 9 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A Question Source: Quad Cities ILT Exam Bank (QDC.LWQ.532012)

Question History: N/A 10 CFR Part 55 Content: 41(b)(10)

SRO Justification: N/A Comments: None Associated objective(s):

SR-ARP-K08 (Freq: LIC=B)

Given a QCARP casualty, STATE the immediate control room actions of the appropriate QCARP.

2.4.11 Knowledge of abnormal condition procedures. (RO=4.0 / SRO=4.2)

Page: 146 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 74 ID: QDC.ILT.16480 Points: 1.00 A room in the Unit 2 Reactor Building was recently surveyed and the following radiological conditions exist:

• General area radiation of 150 mrem/hour.

• Smearable contamination of 1100 dpm/100 cm2 (beta-gamma).

When you arrive at the room, the posted signs are as follows:

• Caution - Radiation Area

• Caution - Contaminated Area What actions (if any) are required/allowed?

A. Continue with assigned work in the area.

B. Do NOT proceed, notify the RP Department that ONLY the Radiation Area posting is NOT correct.

C. Do NOT proceed, notify the RP Department that ONLY the Contaminated Area posting is NOT correct.

D. Do NOT proceed, notify the RP Department that the Radiation Area AND Contaminated Area postings are NOT correct.

Answer: B Answer Explanation:

Personnel entering RCAs are responsible for reading and complying with associated postings and labels.

A Radiation Area is greater than 5 mrem/hour but less than 100 mrem/hour. A High Radiation Area is greater than 100 mrem/hour but less than 1000 mrem/hour. A Contamination Area is greater than 1000 dpm/100 cm2. If the expected conditions are not what is found at the job area, the radiation worker is required to report the incorrect posting to the RP Department.

Distractor 1: Plausible if candidate assumes that the radiation posting is correct for the dose rate given.

Distractor 2: Combination of distractor 1 and 3.

Distractor 3: Plausible if candidate assumes that the contamination posting is incorrect for the given smearable contamination levels.

Reference:

RP-AA-376 Rev 4 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A Question Source: Dresden ILT Exam Bank Page: 147 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Question History: Dresden 2008 ILT NRC Exam 10 CFR Part 55 Content: 41(b)(12)

SRO Justification: N/A Comments: None Associated objective(s):

2.3.07 Ability to comply with radiation work permit requirements during normal or abnormal conditions. (RO=3.5 / SRO=3.6)

NGET Objective link (Refer to Non-Acredited Project for NGET/RWT objectives)

Page: 148 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion) 75 ID: QDC.ILT.16487 Points: 1.00 Unit 1 was operating at 15% power when an automatic scram occurred.

• Multiple control rods failed to insert.

• IRMs are being inserted.

• APRMs are indicating between 1 and 2%.

• Reactor water level dropped to +12 inches and returned to +30 inches.

• Torus temperature is 93°F.

• Torus water level is +1 inch.

• Drywell pressure is 2.8 psig.

• 901-3 D-1, TURB BLDG HI RADIATION annunciator is in alarm Based on these conditions, which QGA(s), if any, are required to be entered directly?

A. No QGA procedure entry is required.

B. QGA 200, Primary Containment Control only.

C. QGA 100 (RPV Control) and QGA 200 (Primary Containment Control) only.

D. QGA 100 (RPV Control), QGA 200 (Primary Containment Control) and QGA 300 (Secondary Containment Control).

Answer: C Answer Explanation:

QGA 100 and 200 are entered on high drywell pressure (2.5 psig).

Distractor 1: Plausible if candidate does not recognize that drywell pressure exceeds the EOP entry condition.

Distractor 2: Plausible if candidate does not recognize that high drywell pressure is an entry condition for QGA 100 and QGA 200.

Distractor 3: Plausible if candidate assumes that Turbine Building high radiation (vice Reactor Building high radiation) is an entry condition for QGA 300.

Reference:

QGA 200 Rev 9 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A Question Source: Quad Cities Exam Bank (QDC.LORTB.16555)

Question History: N/A 10 CFR Part 55 Content: 41(b)(10)

SRO Justification: N/A Comments: None Page: 149 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (RO Portion)

Associated objective(s):

SR-0001-K21 (Freq: LIC=B)

STATE the entry conditions to QGA 200, 'Primary Containment Control.'

2.4.04 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (RO=4.5 /

SRO=4.7)

Page: 150 of 150 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 76 ID: QDC.ILT.16388 Points: 1.00 Unit 1 was in Mode 2 performing a startup when IRM Channel 12 spiked HIGH and immediately returned to its normal value.

The following indications are present on the 901-5 panel:

• A-5, IRM HIGH is slow-flashing

• C-10, CHANNEL A IRM HIGH HIGH OR INOP is slow-flashing

• RPS Channel A scram solenoid GROUP 1 and 4 lights are OUT

• RPS Channel A scram solenoid GROUP 2 and 3 lights are LIT

• ALL RPS Channel B scram solenoid group lights are LIT Which of the following actions are required to be directed FIRST?

A. BYPASS the affected IRM per QCOP 0700-02, IRM Operation.

B. Insert a FULL scram per QCGP 2-3, Reactor Scram.

C. Insert a HALF scram on RPS A per QCOA 0500-01, Partial Scram Actuation.

D. RESET the half scram on RPS A per QCOP 0500-03, Resetting Scrams.

Answer: C Answer Explanation:

Using the available indications, the candidate must determine that a partial half scram has occurred (all 4 group lights should be out on RPS Channel A). During a partial scram actuation, Operator response must be focused on plant safety requirements and not on investigating the RPS system. The proper response is to force all lights in the affected RPS Channel to de-energize, which ensures that all control rods move together if a half scram occurs in the other channel.

Distractor 1: Plausible because this would be the required action if RPS had responded to this event as designed.

Distractor 2: Plausible because a full scram is required for a partial "full" scram.

Distractor 3: Plausible if candidate assumes that the scram should be reset to ensure that a half scram on the other channel does not result in only 2 groups of control rods inserting into the core.

Reference:

QCOA 0500-01 Rev 8 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 1 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 43(b)(5)

Page: 1 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

SRO Justification: Candidate must assess plant conditions and then set priorities by selecting a procedure to implement FIRST based on guidance contained within the AOPs.

Comments: None.

Associated objective(s):

295006.AA2.05 (CFR 41.10, 43.5, 45.13)

Ability to determine and/or interpret the following as they apply to SCRAM : Whether a reactor SCRAM has occurred (RO=4.6 / SRO=4.6)

SR-0500-K26 (Freq: LIC=B)

EVALUATE given key Reactor Protection System parameter indications and/or responses depicting a system specific abnormality/failure and DETERMINE a course of action to correct or mitigate the following abnormal condition(s):

a. Partial half scram

b. Failure to reset

c. Failure to scram

d. Half-scram

e. Loss of one RPS bus

f. Loss of both RPS bus Page: 2 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 77 ID: QDC.ILT.16389 Points: 1.00 A normal plant Shutdown/Cooldown is in progress on Unit 1 with the following conditions:

• RPV pressure is 5 psig and slowly lowering

• No surveillances are in progress

• 1C RHR pump is out-of-service

• 1B RHR and 1A RHRSW pumps are operating in the Shutdown Cooling Mode If the 1B RHR pump TRIPS due to equipment failure, which of the following is the correct action, if any, to be taken in regards to LCO 3.4.7, RHR SDC System - Hot Shutdown?

A. NO action required. LCO 3.4.7 is NOT applicable with the current plant conditions.

B. NO action required. LCO 3.4.7 is met with the remaining OPERABLE equipment.

C. Declare ONE (1) required RHR SDC subsystem inoperable.

D. Declare TWO (2) required RHR SDC subsystems inoperable.

Answer: B Answer Explanation:

To meet the LCO, both RHR pumps (and two RHR service water pumps) in one loop or one RHR pump (and one RHR service water pump) in each of the two loops must be OPERABLE. In this case, the 1A and 1D RHR pumps (and associated RHRSW pumps) remain operable, therefore, TWO subsystems are operable per LCO 3.4.7 basis.

Distractor 1: Plausible if candidate assumes that the applicability for LCO 3.4.7 is Mode 4 (applicability for LCO 3.4.8).

Distractor 2: Plausible if candidate assumes that two RHR pumps constitutes one subsystem.

Distractor 3: Plausible if candidate assumes that two RHR pumps must be operable in each subsystems for operability (similar to RHR-LPCI mode of operation).

Reference:

LCO 3.4.7 Amendment No. 223/218, LCO 3.4.7 Bases Revision 0 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 1 Group: 1 Question Source: Modified from Quad Cities ILT Bank (QDC.ILT.737508)

Question History: N/A 10 CFR Part 55 Content: 43(b)(2)

SRO Justification: The operability determination cannot be made without knowledge of the basis of Tech Spec LCO 3.4.7.

Comments: None.

Page: 3 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Associated objective(s):

295021.2.2.37 Ability to determine operability and/or availability of safety related equipment. (RO=3.6 / SRO=4.6)

S-1000-K33 (Freq: LIC=B)

Discuss the bases for RHR/RHRSW System LCO's.

Page: 4 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 78 ID: QDC.ILT.16390 Points: 1.00 Unit 1 is in a refuel outage with fuel moves in progress when an IRRADIATED fuel bundle is accidentally raised ABOVE the normal up position limit.

Numerous alarms are in, including:

• 901-3 G-16, FUEL POOL CHANNEL A HI RADIATION

• 901-5 D-8, CONTROL ROOM VENT ISOLATED

• 912-5 A-1, RX BLD 1 VENT/EXH FAN TRIP

• 912-5 A-6, STAND BY GAS TREATMENT SYS A TROUBLE Which ONE of the following actions must the SRO prioritize to complete FIRST based on the above annunciators?

A. Evacuate the refuel platform and control refuel floor access per QCFHP 110-7, Irradiated Fuel Above Normal Up Position.

B. Start a Control Room AFU Booster Fan per QCOP 5750-09, Control Room Ventilation System.

C. Restore Reactor Building Ventilation to a normal lineup per QCOP 5750-02, Reactor Building Ventilation System.

D. Verify a Standby Gas Treatment system train is in operation per QCOA 7500-01, Standby Gas Treatment System Auto Start.

Answer: A Answer Explanation:

It is an immediate operator action to evacuate the refuel platform and control access to the refuel floor when an ARM alarms due to irradiated fuel above the normal up position limit. All other distractors contain subsequent actions or actions that are not required.

Therefore, the immediate operator action is the priority.

Distractor 1: Plausible because the AFU is sometimes required to be started in casualty situations following control room ventilation isolation. However, the AFU fan is not required to be started unless a LOCA occurs.

Distractor 2: Plausible because the abnormal procedure for Reactor Building ventilation isolation directs restoration following isolation. However, the isolation is expected for fuel pool high radiation and SBGT will maintain secondary containment d/p.

Distractor 3: Plausible because the abnormal procedure has the operator verify SBGT system train operation, however, it is not an immediate operator action.

Reference:

QCFHP 0110-7 Rev 0 Reference provided during examination: None.

Cognitive level: High Level (RO/SRO): SRO Tier: 1 Group: 1 Question Source: New Page: 5 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Question History: N/A 10 CFR Part 55 Content: 43(b)(7)

SRO Justification: The immediate operator action tested is contained in a fuel handling abnormal procedure (QCFHP), NOT an operations abnormal procedure (QCOA).

Therefore, the candidate must have knowledge of fuel handling procedures to successfully answer this question.

Comments: None Associated objective(s):

SRLF-805-K16 (Freq: LIC=B NF=B)

Given a Refueling Operations related casualty, STATE the immediate operator actions of the applicable abnormal procedure.

295023.2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (RO=4.1 / SRO=4.3)

Page: 6 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 79 ID: QDC.ILT.16418 Points: 1.00 (Use your provided reference to answer this question.)

Unit 2 is operating at rated power when the Unit 2 125 VDC Station Battery is determined to be INOPERABLE.

Complete the following statement per Technical Specifications:

If the Unit 2 125 VDC electrical loads are transferred to the Unit 2 125 VDC Alternate Battery, Unit 2 operation at rated power...

A. CAN continue indefinitely.

B. can NOT continue indefinitely because the location of the alternate battery makes it susceptible to single failure.

C. can NOT continue indefinitely because the alternate battery does NOT have the capacity to carry additional loads during a DBA for a period of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

D. can NOT continue indefinitely because the alternate battery charger does NOT have the capacity to carry normal loads and maintain the alternate battery on a float charge.

Answer: B Answer Explanation:

When the station 125 VDC battery is inoperable (whether due to loss of voltage or testing/maintenance), placement of the operable alternate 125 VDC electrical power subsystem in-service helps ensure that the design basis can be met. However, the design configuration of the alternate battery (location) is susceptible to failure and hence, is not as reliable as the normal battery. Therefore, only a limited time of operation is allowed in this condition.

Distractor 1: Plausible because the alternate battery has the capacity to carry required design basis loading.

Distractor 2: Plausible because part of the design basis of the DC sources is to carry the loading required for a DBA on one unit, and safe shutdown of the other unit for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

Incorrect because the alternate battery capacity is adequate.

Distractor 3: Plausible because the alternate battery charger is smaller than the two normal chargers.

Reference:

TS B 3.8.4 Rev 40, TS LCO 3.8.4 Amendment No. 245/240 Reference provided during examination: TS LCO 3.8.4 with statement, applicability and surveillance requirements removed.

Cognitive level: Memory Level (RO/SRO): SRO Tier: 1 Group: 1 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.05504)

Question History: N/A Page: 7 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 10 CFR Part 55 Content: 43(b)(2)

SRO Justification: Candidate must have knowledge of Technical Specifications and their bases.

Comments: None Associated objective(s):

S-6900-K33 (Freq: LIC=I)

DISCUSS the bases for Station DC Electrical Systems related Tech Spec LCO's.

295004.2.1.07 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (RO

= 4.4 / SRO = 4.7)

Page: 8 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 80 ID: QDC.ILT.16393 Points: 1.00 Unit 2 was at rated power when an accident occurred resulting in the following plant conditions:

• Torus water level is 12.0 feet and steady

• Torus water temperature is 190°F and steady

• Reactor pressure is 400 psig and steady Complete the following statements:

Given the above plant conditions, the Heat Capacity Temperature Limit (HCTL) __(1)__

exceeded.

Per the EOP basis, maintaining plant conditions within the HCTL will __(2)__.

A. (1) IS (2) prevent ECCS pump damage from inadequate net positive suction head B. (1) IS NOT (2) prevent ECCS pump damage from inadequate net positive suction head C. (1) IS (2) ensure Torus pressure remains below the Primary Containment Pressure Limit (PCPL) during a RPV blowdown D. (1) IS NOT (2) ensure Torus pressure remains below the Primary Containment Pressure Limit (PCPL) during a RPV blowdown Answer: C Page: 9 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Answer Explanation:

The HCTL is the highest torus temperature from which a RPV blowdown will not raise (a) torus temperature above the torus design temperature and (b) torus pressure above the PCPL, while the rate of energy transfer from the RPV to the primary containment is greater than the capacity of the containment vent.

Distractor 1: Plausible because pump damage is a concern when suppression pool temperature is high (displayed as a caution in the EOP torus temperature leg). However, a separate graph is used to determine margin to pump damage from inadequate NPSH.

Distractor 2: Combination of distractor 1 and 2.

Distractor 3: Plausible if candidate determines that operation must be above the line, similar to the BFPL curves for the RPV.

Reference:

L-QGADET Rev 8 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 1 Group: 1 Question Source: Modified from Quad Cities LORT question (QDC.LORTB.0312)

Question History: N/A 10 CFR Part 55 Content: Unique to the SRO position SRO Justification: Candidate must perform actions that are unique to the SRO position (reference facility objective).

Comments: None Associated objective(s):

295026.2.4.18 Knowledge of the specific bases for EOPs. (RO=3.3 / SRO=4.0)

S-0001-K12a (Freq: LIC=B)

EVALUATE given system/plant parameters and the following QGA curves/tables, DETERMINE if any QGA related limits have been exceeded:

a. QGA Detail A, RPV Water Level Instruments

1. Figure B, RPV Saturation Curve

2. Table C, RPV Level Instrument Criteria

b. QGA Figure D, Primary Containment Pressure Limit

c. QGA Table J, Minimum Steam Cooling Pressure

d. QGA Figure K, Drywell Spray Initiation Limit

e. QGA Figure L, Pressure Suppression Pressure

f. QGA Figure M, Heat Capacity Limit

g. QGA Table S, Reactor Building Area Temperatures

h. QGA Table T, Reactor Building Area Radiation Levels

i. QGA Table U, Reactor Building Area Water Levels Page: 10 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 81 ID: QDC.ILT.16394 Points: 1.00 (Use your provided references to answer this question.)

A LOCA with a loss of offsite power (LOOP) has occurred on Unit 2. Current plant parameters are:

• RPV pressure 50 psig

• RPV water level -250 inches and lowering

• Drywell pressure 21 psig

• Torus pressure 19 psig

• Drywell H2 concentration 3%

• Drywell O2 concentration 1%

• Torus H2 concentration 2%

• Torus O2 concentration 1%

• All available systems are injecting into the RPV.

Which of the actions listed below has the GREATEST priority?

A. Exit all QGAs AND enter all SAMGs provided the TSC is prepared to provide SAMG decision-making.

B. Enter QGA 200-5, Hydrogen Control, and vent and purge the Drywell if the offsite release rate is expected to stay below the LCO limit.

C. Enter QCOP 1000-30, Post-Accident RHR Operation, and initiate Drywell Sprays.

D. Enter QCOP 1600-13, Post Accident Venting of the Primary Containment, and vent the Torus even if release rate limits are exceeded.

Answer: A Answer Explanation:

The SAMGs are entered when "adequate core cooling" cannot be established. QGA 100 and QGA 101 specifically direct entry into the SAMGs and exit from ALL QGAs when core cooling is lost and the TSC is prepared to provide SAMG decision-making.

Any direction of QCOPs or QCOAs from the QGAs are also overridden by SAMG direction.

Distractor 1: QGA directed action from QGA 200 on detection of Hydrogen in the containment. Is not a priority over SAMG actions.

Distractor 2: Per QGA 200 , flow cannot be diverted from the core if adequate core cooling is not established.

Distractor 3: PCPL has not been violated, so the Torus cannot be vented irrespective of the release rate.

Reference:

QGA 100 Rev. 9 Reference provided during examination: QGA 100, 200 and 200-5 (delete entry conditions, Torus temperature leg information, Low Torus level actions for 11 ft and Caution statement in Detail A)

Page: 11 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Cognitive level: High Level (RO/SRO): SRO Tier: 1 Group: 1 Question Source: Quad Cities Bank (QDC.ILT.15565)

Question History: 2009 ILT NRC Exam 10 CFR Part 55 Content: 43(b)(5)

SRO Justification: Requires candidate to assess plant conditions and, using the EOP flowcharts, select and supporting procedure and action to mitigate with. Also is supported by a SRO only facility objective.

Comments: None Associated objective(s):

S-0001-K18 (Freq: LIC=B)

Given QGA 100, RPV Control, and various conditions, EVALUATE the conditions and DESCRIBE how to proceed through the flowchart including transitions within QGA 100, to other QGA procedures, station operating procedures, or SAMGs.

295031.EA2.04 (CFR 41.10, 43.5, 45.13)

Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling. (RO=4.6 / SRO=4.8)

Page: 12 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 82 ID: QDC.ILT.16395 Points: 1.00 Unit 2 is at 100% power when a transient occurs, resulting in a HIGH reactor pressure condition.

Which of the following conditions, if any, will require the Shift Manager to declare an ALERT due to meeting the threshold values of MA3? If neither condition requires a declaration, why?

Condition 1: RPV pressure reaches 1103 psig and reactor shutdown is achieved by automatic scram solenoid pilot valve operation.

Condition 2: RPV pressure reaches 1257 psig and reactor shutdown is achieved by automatic ARI initiation.

(Note: Consider each condition separately.)

A. Condition 2 only B. Condition 1 and 2 C. NEITHER because MA3 is NOT applicable for the initial operating condition.

D. NEITHER because the reactor was automatically shutdown in both conditions.

Answer: A Answer Explanation:

The candidate must determine:

1) if reactor pressure is above the RPS scram setpoint.

2) if ARI and scram solenoid pilot valves are normal or alternate means of rod insertion.

3) if MA3 is applicable if the reactor is shutdown and in Mode 3.

The UFSAR analytical value for RPS scram on high RPV pressure is 1060 psig (actual is 1024 psig). Therefore, anything less than 1024 psig does not meet threshold value one (1).

Page: 13 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

The second condition of this EAL indicates a failure of the automatic RPS scram function to rapidly insert a sufficient number of control rods to achieve reactor shutdown. The Alternate Rod Insertion (ARI) system provides an automatic and alternate method of completing the scram function. This backup, however, inserts control rods at a much slower rate than the automatic RPS scram function. For the purpose of emergency classification at the Alert level, reactor shutdown achieved by ARI initiation does not constitute a successful RPS automatic scram.

Since the ATWS condition (for Condition 2) was initially present with the plant in Mode 1, MA3 is applicable (even if ARI places the plant in Mode 3).

Distractor 1: Plausible if candidate does not recognize that reactor shutdown from automatic scram solenoid pilot valve operation is the normal method of control rod insertion.

Distractor 2: Plausible if candidate incorrectly applies the applicability of MA3 to the initial condition of the plant.

Distractor 3: Plausible if candidate assumes that declaring an alert is not required because the reactor was automatically shutdown in both conditions.

Reference:

EP-AA-1006 Rev 29 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 1 Group: 1 Question Source: Modified from Quad ILT Bank (QDC.ILT.760618)

Question History: N/A 10 CFR Part 55 Content: 41(b)(10)

SRO Justification: Unique to SRO position (reference facility objective)

Comments: None Associated objective(s):

295037.EA2.06 (CFR 41.10, 43.5, 45.13)

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

Reactor pressure (RO=4.0 / SRO=4.1)

S-0303-K70 (Freq: LIC=B ILT=NA) Given an ATWS/ARI System operating mode and various plant conditions and a copy of EP-AA-111 and EP-AA-1006, CLASSIFY the event/abnormal condition including correct EALs and PARs in accordance with EP-AA-111 and EP-AA-1006.

Page: 14 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 83 ID: QDC.ILT.16397 Points: 1.00 (Use your provided references to answer this question.)

Unit 2 is in Day 2 of a refuel outage with the following plant conditions:

• 2A CRD pump is RUNNING

• CRD charging water header is ISOLATED

• All 177 HCUs are in the process of being hydraulically ISOLATED

• Fuel is NOT being moved within the reactor vessel With the HCU isolations in progress, control rods D-7, E-6 and F-6 drift out to positions 12, 18 and 22 respectively and STOP.

One (1) minute later, the following indications are present on the 902-5 panel:

If operator action is then taken and ALL control rods are inserted to position 00, what is the HIGHEST Reactivity Management significance level that must be assigned to this event?

A. Significance Level 1 B. Significance Level 2 C. Significance Level 3 D. Significance Level 4 Answer: A Answer Explanation:

This question is modeled after SEN 264 and a significant event at Dresden station in 2008 (OE28297).

Page: 15 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Control rods D-7, E-6 and F-6 are near SRM 24. Per QCFHP 0110-02, true criticality is indicated by a sustained increase in count rate, over 15 to 20 seconds, of the SRM closest to the control rod being moved.

Therefore, an inadvertent criticality has occurred during refueling operations. That is a Level 1 significant event.

Distractor 1: Plausible if candidate does not determine that the reactor is critical because indications of criticality are only associated with SRM 24. A classification to Level 2 would be based on entry into Tech Spec actions for loss of Shutdown Margin.

Distractor 2: Plausible if candidate does not fully evaluate all levels of significance and determines a Level 3 based on mispositioned control rod due to degraded equipment.

Distractor 3: Plausible if candidate does not fully evaluate all levels of significance and determines a Level 4 based on an observable change in reactor power of < 0.5% RTP caused by personnel error.

Reference:

OP-AA-300-1540 Rev 6, QCFHP 0110-02 Rev 4 Reference provided during examination: OP-AA-300-1540 Attachment 1 Rev 6 Cognitive level: High Level (RO/SRO): SRO Tier: 1 Group: 2 Question Source: Modified from Quad Cities ILT Exam Bank (QDC.ILT.15502)

Question History: Original version used on Quad 2009 ILT NRC Exam 10 CFR Part 55 Content: 43(b)(6)

SRO Justification:

Comments: None Associated objective(s):

SRLF-PGRM-K4 (Freq: LIC=A NF=A)

Given various reactor conditions and evolutions in progress, IDENTIFY the activities that impact core reactivity and/or have reactivity management implications.

295014.AA2.02 (CFR 41.10, 43.5, 45.13)

Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION: Reactor period (RO=3.9 / SRO=3.9)

Page: 16 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 84 ID: QDC.ILT.16398 Points: 1.00 (Use your provided references to answer this question.)

Unit 2 was operating at rated power when an accident occurred involving a LEAK from the north scram discharge volume (SDV). Current plant conditions are as follows:

• Reactor has been SCRAMMED for 45 minutes

• ALL attempts to reset the scram have FAILED

• Drywell Radiation Monitor 'A' indicates 550 R/hr and rising

• Drywell Radiation Monitor 'B' indicates 600 R/hr and rising

• North CRD Module Area ARM indicates 3336 mr/hr and rising

• South CRD Module Area ARM indicates 2810 mr/hr and rising Which of the following is the HIGHEST emergency action level of classification for the above plant conditions?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer: C Answer Explanation:

Site Area Emergency (FS1) must be declared based on meeting the threshold values of RC4 (for potential loss) and CT6.

Based on plant conditions, candidate must determine that the North CRD Module Area is on the first floor of the reactor building (595' elevation), and therefore a radiation level of QGA 300 has exceeded both normal and maximum safe levels.

Candidate must also determine that a leak from the SDV with a failure to reset the scram constitutes an unisolable primary system leak.

Distractor 1: Plausible if candidate determines EAL classification by reading the Fission Product Barrier Matrix chart from right-to-left. Candidate would arrive at an Unusual Event based on CT5 or CT6 and then make the declaration. Operators are specifically trained to classify by reading right-to-left to prevent under classification of events.

Distractor 2: Plausible if candidate only references the Abnormal Radiation Levels /

Radiological Effluent matrix and classifies the event due to meeting the threshold values of RA3. Also plausible if candidate determines that threshold value of RC4 (for potential loss) is met but does not recognize that CT6 is also met with the same plant conditions.

Distractor 3: Plausible if candidate misinterprets Table F1, assumes a loss of the fuel clad barrier and declares a General Emergency based on FG1.

Reference:

EP-AA-1006 Rev 29, QGA 300 Rev 11, QCOP 1800-01 Rev 12 Reference provided during examination: EP-AA-1006 Rev 29 pages QC 3-11 thru 3-12, QGA 300 Rev 1 (with entry conditions, QGA 400 and Detail A deleted)

Page: 17 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Cognitive level: High Level (RO/SRO): SRO Tier: 1 Group: 2 Question Source: New Question History: N/A 10 CFR Part 55 Content: Unique to SRO position SRO Justification: Determination of emergency classification is a function of the SRO position (see facility objective).

Comments: None Associated objective(s):

S-0001-K70d (Freq: LIC=B ILT=NA) Given a various plant conditions related to QGA 300, RPV Control, and a copy of EP-AA-111 and EP-AA-1006, CLASSIFY the event/abnormal condition including correct EALs and PARs in accordance with EP-AA-111 and EP-AA-1006.

295033.2.4.41 Knowledge of the emergency action level thresholds and classifications.

(RO=2.9 / SRO=4.6)

Page: 18 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 85 ID: QDC.ILT.16507 Points: 1.00 (Use your provided reference to answer this question.)

The following conditions exist on Unit 2:

• Reactor Mode Switch is in REFUEL.

• Mode 5 was entered 8 DAYS ago.

• Fuel moves are in progress within the Reactor Pressure Vessel (RPV)

• Control Rod Drive (CRD) removal from under the vessel is in progress. This activity has been screened as an Operation with the Potential to Drain the Reactor Vessel (OPDRV).

A Reactor Building Ventilation exhaust fan TRIPS, resulting in the indication BELOW.

What action, if any, is required per TS LCO 3.6.4.1, Secondary Containment and why?

A. NO action is required because the secondary containment is OPERABLE.

B. NO action is required because current plant conditions do NOT require the secondary containment to be operable.

C. IMMEDIATELY initiate action to suspend CRD removal because the secondary containment is INOPERABLE.

D. IMMEDIATELY suspend core alterations because the secondary containment is INOPERABLE.

Answer: C Page: 19 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Answer Explanation:

Per SR 3.6.4.1.1 requires secondary containment d/p to be > 0.01 inch of vacuum water gauge. The given indication does not meet SR 3.6.4.1.1, therefore the secondary containment is INOPERABLE.

With OPDRVs in progress, immediate action must be taken to suspend OPDRVs.

Distractor 1: Plausible if candidate interprets the > 0.10 inch vacuum water gauge to mean "greater than" in the positive direction.

Distractor 2: Plausible if candidate assumes that the secondary containment is not required to be operable given the current plant conditions.

Distractor 3: Plausible if candidate assumes that recently (< 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />) irradiated fuel is being moved within the secondary containment.

Reference:

TS LCO 3.6.4.1 Amendment No. 245/240 Reference provided during examination: TS 3.6.4.1 and TS 3.6.4.1 Surveillance Requirements with the following statements deleted:

TS 3.6.4.1 LCO and Applicability statements, Condition C, Required Action C.2, and Condition C Completion Times.

Cognitive level: High Level (RO/SRO): SRO Tier: 1 Group: 2 Question Source: New Question History: N/A 10 CFR Part 55 Content: 55.43 (2)

SRO Justification: The candidate must determine operability of a system and the required actions in accordance with the rules of application requirements of TS Section 1.

Comments: None.

Associated objective(s):

295035.EA2.01 (CFR 41.8 to 41.10)

Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Secondary containment pressure:

Plant-Specific (RO=3.8 / SRO=3.9)

SR-1601-K32 (Freq: LIC=B)

Given Containment Systems operability status OR key parameter indications, various plant conditions and a copy of Tech Specs, DETERMINE Tech Spec compliance and required actions, if any.

Page: 20 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 86 ID: QDC.ILT.16414 Points: 1.00 The Unit 2 HPCI system is running in its test lineup for a periodic surveillance circulating water to the CCSTs.

Complete the following two statements:

If the HPCI turbine TRIPS on high turbine exhaust pressure, MO 2-2301-14 (MIN FLOW BYP VLV) will __(1)__.

If a HIGH torus water level condition is present after the completion of the surveillance, the PREFERRED (normal) method of lowering torus water level is using __(2)__.

A. (1) OPEN, draining CCST water to the Torus (2) QCOP 1000-12, Torus Water Transfer to Main Condenser B. (1) OPEN, draining CCST water to the Torus (2) QCOP 1000-18, Torus Water Transfer to Floor Drain Collector Tank C. (1) remain CLOSED, maintaining CCST level steady (2) QCOP 1000-18, Torus Water Transfer to Floor Drain Collector Tank D. (1) remain CLOSED, maintaining CCST level steady (2) QCOP 1000-12, Torus Water Transfer to Main Condenser Answer: B Answer Explanation:

When the MO 1-2301-3 (HPCI Turbine Steam Supply) is open, the MO 1-2301-14 (Min Flow) operates as a minimum flow valve. A turbine trip does not close the HPCI 3 valve, therefore the HPCI 14 valve will be open following a turbine trip.

If Torus level is high, transfer of Torus water to the Floor Drain Collector Tank is preferred due to high Torus water conductivity.

Distractor 1: Plausible because torus water can be transferred to the main condenser.

Distractor 2: Plausible if candidate assumes that the minimum flow valve closes on a HPCI trip (e.g. similar to the Core Spray min flow valve closing on a Core Spray pump trip).

Distractor 3: Combination of distractor 1 and 2.

Reference:

QCOP 1600-12 Rev 14, LN-2300 Rev 17 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: 43(b)(5)

Page: 21 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

SRO Justification: Candidate must select appropriate procedure to lower Torus water level based on concerns with high conductivity and subsequent plant operation.

Comments: None Associated objective(s):

206000.A2.01 (CFR 41.5 / 45.6)

Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Turbine trips: BWR-2,3,4 (RO=4.0 / SRO=4.0)

SR-2300-K20 (Freq: LIC=B)

Given a HPCI System operating mode and various plant conditions, EVALUATE the following HPCI System indications/responses and DETERMINE if the indication/

response is expected and normal.

a. MOV and AOV valve positions

b. Turbine inlet and exhaust steam pressures

c. Turbine speed

d. HPCI booster pump suction

e. HPCI main pump discharge pressure

f. HPCI flow rate

g. Turbine oil tank level

h. Turbine lubrication and control oil pressures, temperatures

i. Turning gear engaged/running indications

j. Zero speed indicator light

k. Gland seal condensate (hotwell) pump and leakoff (exhaust) blower run indications

l. Gland seal cooling water pump run indications

m. Auxiliary oil pump and emergency bearing oil pump run indications

n. Gland seal condenser hotwell level

o. Isolation trip channel power supply indications

p. HPCI area temperature and radiation levels

q. Local MGU power indications

r. MSC and MGU position indication Page: 22 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 87 ID: QDC.ILT.16415 Points: 1.00 Unit 1 was at 75% power when there was a complete LOSS of off-site power.

• All rods are fully inserted in the core.

• Reactor Mode switch is in SHUTDOWN.

• The HIGHEST reactor pressure reached during the transient was 1065 psig.

• Reactor pressure is currently 1050 psig and rising at 2 psig per minute.

• QGA 100, RPV Control has been entered but NO further operator action has been taken.

Complete the following statements regarding the execution of the QGA 100 pressure leg:

The correct ANSWER to the decision diamond question "Any ADS Valve Cycling?" is __(1)__.

Given the current plant conditions, the Unit Supervisor must direct the Operator to maintain reactor pressure 800 - 1000 psig using __(2)__.

A. (1) NO (2) main turbine bypass valves per QCGP 2-3, Reactor Scram B. (1) NO (2) ADS valves per QCOP 0203-01, Reactor Pressure Control Using Manual Relief Valve Actuation C. (1) YES (2) ADS valves per QCOP 0203-01, Reactor Pressure Control Using Manual Relief Valve Actuation D. (1) YES (2) main turbine bypass valves per QCGP 2-3, Reactor Scram Answer: B Page: 23 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Answer Explanation:

Reactor pressure of 1065 psig is high, but below the lift setpoint of the low-set relief valves (1115 psig). Therefore, no ADS valves are expected to by cycling.

With a loss of off-site power, main turbine bypass valves are NOT available (MSIVs are shut and power is lost to the EHC pumps). Therefore, alternate pressure control systems must be used. In this case, the ADS valves will serve as pressure control.

Distractor 1: Plausible if candidate does not recognize that the preferred pressure control system (main turbine bypass valves) is not available.

Distractor 2: Plausible if candidate assumes the high reactor pressure condition will actuate the low-set relief valves.

Distractor 3: Combination of distractor 1 and 2.

Reference:

QGA 100 Rev 9, LIC-0203 Rev 16, QOM 1-6700-T03 Rev 4, QOM 1-6700-T01 Rev 6 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: Unique to the SRO position SRO Justification: Candidate must assess plant conditions and select appropriate action within the EOPs (see facility objective).

Comments: None Associated objective(s):

239002.A2.06 (CFR 41.5 / 45.6)

Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Reactor high pressure (RO=4.1 / SRO=4.3)

S-0001-K18 (Freq: LIC=B)

Given QGA 100, RPV Control, and various conditions, EVALUATE the conditions and DESCRIBE how to proceed through the flowchart including transitions within QGA 100, to other QGA procedures, station operating procedures, or SAMGs.

Page: 24 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 88 ID: QDC.ILT.16400 Points: 1.00 Unit 1 was at rated power when a transient occurred resulting in entry into QGA 100, RPV CONTROL.

• Digital Feedwater Level Control (DFWLC) is attempting to control RPV water level.

Which of the following conditions will require the Unit Supervisor to declare RPV water level UNKNOWN and enter QGA 500-4, RPV FLOODING?

Condition 1: RPV pressure is 100 psig, drywell temperature is 250°F and ALL Detail C instruments are DOWNSCALE.

Condition 2: RPV pressure is 50 psig, drywell temperature is 325°F and both Medium Range instruments are reading +30 inches and STEADY.

Condition 3: RPV pressure is 40 psig, drywell temperature is 340°F and ALL Detail C instruments are UPSCALE.

(Note: Consider each condition separately.)

A. Condition 3 only B. Conditions 1 and 3 only C. Conditions 2 and 3 only D. Conditions 1, 2 and 3 Answer: A Page: 25 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Answer Explanation:

The operational implication of Detail A, RPV Water Level Instruments, within the EOPs is being tested. The candidate must analyze each condition and determine which one(s) have the operational implication of "RPV water level unknown."

Determining RPV level can be accomplished by three methods:

1) Directly reading the instruments.

2) Utilizing level correction for the instrument per Attachment A of QCAP 0200-10 to obtain a reading.

3) The RPV level can be considered KNOWN when level is below the usable range per Figure C if there are no other changes to the status of the instruments, such as saturation and reference leg flashing or loss of power. Simply by dropping below the usable range per Figure C the instruments are not usable for a definitive direct reading. However, there is a reasonable assurance the instruments will respond properly when level is restored if nothing has caused the instruments to be considered unreliable.

The only condition that does not meet this guidance is Condition 3 because plant conditions are within the saturation portion of Detail B and all instruments are upscale (which is what happens when reference leg flashing occurs). A caution in Detail A states:

RPV water level instruments may be unreliable due to boiling in the instrument runs when drywell or reactor building temperature is above Fig B, RPV Saturation Temperature.

Distractor 1: Plausible if candidate assumes that RPV level must be considered unknown when all instruments are downscale and outside of their QGA usable range.

Distractor 2: Plausible if candidate assumes that the caution in Detail A states that all instruments must (vice may) be considered unreliable when saturation conditions are present for the instrument reference legs.

Distractor 3: Combination of distractor 2 and 3.

Reference:

QGA 100 Rev 9, QCAP 0200-10 Rev 41 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: Unique to the SRO position.

SRO Justification: Candidate must have knowledge of Detail A use and when to transition to QGA 500-4 (see facility objective).

Comments: None Page: 26 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Associated objective(s):

259002.2.4.20 Knowledge of operational implications of EOP warnings, cautions, and notes. (RO=3.8 / SRO=4.3)

S-0001-K12a (Freq: LIC=B)

EVALUATE given system/plant parameters and the following QGA curves/tables, DETERMINE if any QGA related limits have been exceeded:

a. QGA Detail A, RPV Water Level Instruments

1. Figure B, RPV Saturation Curve

2. Table C, RPV Level Instrument Criteria

b. QGA Figure D, Primary Containment Pressure Limit

c. QGA Table J, Minimum Steam Cooling Pressure

d. QGA Figure K, Drywell Spray Initiation Limit

e. QGA Figure L, Pressure Suppression Pressure

f. QGA Figure M, Heat Capacity Limit

g. QGA Table S, Reactor Building Area Temperatures

h. QGA Table T, Reactor Building Area Radiation Levels

i. QGA Table U, Reactor Building Area Water Levels Page: 27 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 89 ID: QDC.ILT.16413 Points: 1.00 A startup is in progress on Unit 1 with the following plant conditions:

• Reactor Mode switch in STARTUP

• IRMs are on Range 8

• Reactor pressure is 95 psig Which of the following annunciators, if received with the current plant conditions, must be addressed FIRST if the Unit Supervisor places a priority on ensuring the OPERABILITY of Technical Specification required equipment?

(Note: Consider each annunciator separately.)

A. 901-5 A-7, RBM HIGH OR INOP B. 901-3 G-14, AUTO BLOWDOWN INHIBIT C. 901-5 D-8, CONTROL ROOM VENT ISOLATED D. 901-3 C-3, CORE SPRAY PUMP AREA CLR FAN TRIP Answer: D Answer Explanation:

901-3 C-3 annunciator indicates that a breaker for the room cooler in the 'A' or 'B' Core Spray pump room is tripped. If the cooler cannot be restored, the Core Spray and/or RCIC subsystem will be required to be declared inoperable.

Distractor 1: Plausible if candidate assumes that the RBM is required per Technical Specifications during a startup. Incorrect because reactor power is < 30% RTP.

Distractor 2: Plausible because this alarm indicates that the ADS function of the relief valves will not work. However, the ADS function is not required due to steam dome pressure <150 psig. The relief, safety and manual functions of the valves remain operable with this alarm in.

Distractor 3: Plausible if candidate assumes that the 'B' train of control room ventilation is impacted by an isolation. A control room isolation will prevent the 'A' (non-safety related) train from running in the normal mode, but will not prevent the 'B' train from satisfying its safety function. Also plausible because reactor building ventilation isolation results in the SRO declaring a Tech Spec required equipment inoperable (rad monitors).

Reference:

QCAN 901(2)-3 F- Rev 4 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 2 Group: 1 Question Source: New Question History: N/A 10 CFR Part 55 Content: Unique to the SRO position Page: 28 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

SRO Justification: Candidate must be able to determine operability of Tech Spec required equipment (see facility objective).

Comments: None Associated objective(s):

209001.2.4.45 Ability to prioritize and interpret the significance of each annunciator or alarm. (RO=4.1 / SRO=4.3)

S-OPDT-K06 (Freq: LIC=A) Given a degraded or nonconforming condition that may impact the operability of a specific SSC described in Tech Specs, using P&ID/C&IDs, E-prints and Tech Specs, if necessary, PERFORM an immediate Operability Determination and DETERMINE if the SSC meets Tech Spec operability requirements in accordance with the Operability Determination procedures, OP-AA-108-115 and OP-AA-108-115-1002.

Page: 29 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 90 ID: QDC.ILT.16401 Points: 1.00 Unit 1 is shutdown and in Mode 3, Unit 2 is at rated power.

• Emergent maintenance requires MCC 18-2 to be DE-ENERGIZED.

If MCC 18-2 is removed from service, determine the OPERABILITY of affected AC distribution subsystems in accordance with LCO 3.8.7, Distribution Systems - Operating, for BOTH Units.

A. ALL required AC subsystems are OPERABLE for both Unit 1 and Unit 2.

B. One (1) required AC subsystem is INOPERABLE for both Unit 1 and Unit 2.

C. One (1) required AC subsystem is INOPERABLE for Unit 1. All required AC subsystems are OPERABLE for Unit 2.

D. ALL required AC subsystems are OPERABLE for Unit 1. One (1) required AC subsystem is INOPERABLE for Unit 2.

Answer: B Answer Explanation:

The Essential Service (ESS) Bus is required as an AC safety bus subsystem.

De-energizing MCC 18-2 will not de-energize the ESS bus (UPS is still powering bus).

However, TS B 3.8.7 states that the 120 VAC ESS Bus must be capable of being energized from Bus 18-2 (28-2).

Operability requirements of the opposite unit's Division 1 AC electrical power distribution subsystem requires operability of the Unit 1 ESS Bus.

Distractor 1: Plausible if candidate does not know that the ESS bus must be capable of being energized from MCC 18-2 to be considered operable in Modes 1, 2 and 3.

Distractor 2: Plausible if candidate does not know that the opposite unit ESS bus must be operable to meet the requirements of TS LCO 3.8.7.

Distractor 3: Plausible if candidate assumes that LCO 3.8.7 is only applicable in Modes 1 and 2 (i.e. when operating).

Reference:

TS LCO 3.8.7 Amendment No. 245/240, TS B 3.8.7 Rev 40 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 2 Group: 1 Question Source: Modified from Quad Cities Bank (QDC.LN.604715)

Question History: N/A 10 CFR Part 55 Content: 43(b)(2)

SRO Justification: Candidate must have knowledge of Technical Specification bases and determine operability based on plant equipment availability.

Page: 30 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Comments: None Associated objective(s):

262002.2.2.36 Ability to analyze the affect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. (RO=3.1 / SRO=4.2)

S-6500-K33 (Freq: LIC=I) DISCUSS the bases for 4KV / 480 VAC Distribution Systems related Tech Spec LCO's.

Page: 31 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 91 ID: QDC.ILT.16402 Points: 1.00 Unit 1 is in Mode 2 performing a startup with reactor power at 3%.

• Control rod G-6 has just been declared INOPERABLE and must be fully inserted into the core.

• RWM Mode switch is in NORMAL.

Using the information on the Rod Worth Minimizer (RWM) display below, complete the following statements:

If control rod G-6 is selected on the rod select matrix, an INSERT BLOCK will be applied __(1)__.

If the Unit Supervisor directs the Reactor Operator to BYPASS the RWM, LCO 3.3.2.1 (Control Rod Block Instrumentation) must be entered because the RWM ensures that the initial conditions of the __(2)__ analysis are NOT violated.

A. (1) immediately (2) single control rod withdrawal error (RWE)

B. (1) immediately (2) control rod drop accident (CRDA)

C. (1) ONLY after the rod is inserted to position 46 (2) control rod drop accident (CRDA)

D. (1) ONLY after the rod is inserted to position 46 (2) single control rod withdrawal error (RWE)

Answer: B Answer Explanation:

Page: 32 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Candidate must recognize that control rod G-6 is out-of-sequence (not green in color).

Candidate must also recognize that control rod G-6 is in its expected position (white in color).

Any out-of-sequence rod selected will cause an insert and withdraw block to be applied for the selected rod with one exception: insertion/withdrawal of an out-of-sequence rod is allowed in order to correct a withdraw/insert error.

The RWM is required by Technical Specifications when < 10% RTP to satisfy the initial conditions of the Control Rod Drop Accident (CRDA).

Distractor 1: Plausible if candidate assumes that the RWM satisfies the initial conditions of the single rod withdraw error event. This is the basis of the Rod Block Monitor and is a frequent misconception.

Distractor 2: Plausible if candidate does not apply the concept of "selection error" for the RWM and assumes that the sequence must be violated before the RWM enforces a rod block.

Distractor 3: Combination of distractor 1 and 2.

Reference:

QCOP 0207-01 Rev 18, TS B 3.3.2.1 Rev 28 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 2 Group: 2 Question Source: New Question History: N/A 10 CFR Part 55 Content: 43(b)(2)

SRO Justification: Question requires candidate to have knowledge of TS LCO 3.3.2.1 bases for the RWM.

Comments: None Associated objective(s):

201006.2.2.44 Ability to interpret control room indications to verify the status and operation of system, and understand how operator actions and directives affect plant and system conditions. (RO=4.2 / SRO=4.4)

S-0207-K33 (Freq: LIC=I)

DISCUSS the bases for Rod Worth Minimizer Tech Spec LCOs.

Page: 33 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 92 ID: QDC.ILT.16410 Points: 1.00 Unit 2 is operating at 65% power with all OPRMs inoperable.

• A TRIP of the 2A Reactor Recirc pump occurs.

Complete the following two statements:

Following the pump trip, INDICATED reactor water level on LI 2-0640-29A/B "Narrow Range" instruments will INITIALLY __(1)__.

If plant operation stabilizes within Instability Region I, the Unit Supervisor is required to enter

__(2)__ to exit the instability region.

A. (1) lower (2) QCGP 2-3, Reactor Scram, and initiate a reactor scram B. (1) lower (2) QCOA 0400-02, Core Instabilities, and insert control rods C. (1) rise (2) QCOA 0400-02, Core Instabilities, and insert control rods D. (1) rise (2) QCGP 2-3, Reactor Scram, and initiate a reactor scram Answer: D Page: 34 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Answer Explanation:

Immediately following a reactor recirc pump trip, core flow lowers and void fraction within the core increases. This results in a higher core d/p, therefore reactor water level increases until DFWLC responds.

With the OPRM system inoperable, plant operation is not allowed within Instability Region I.

Distractor 1: Plausible if candidate assumes that water level lowers because initially core flow is lower with steam flow constant.

Distractor 2: Combination of distractor 1 and 3.

Distractor 3: Plausible because if the OPRMs were operable, plant operation can continue if control rods are inserted to exit Instability Region I.

Reference:

QCOA 0400-02 Rev 20 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 2 Group: 2 Question Source: Modified from Hatch ILT Exam Bank Question History: Hatch 2009 ILT NRC Exam 10 CFR Part 55 Content: 43(b)(5)

SRO Justification: Candidate must assess plant conditions and select appropriate procedure during abnormal operation.

Comments: None Associated objective(s):

SR-0201-K22 (Freq: LIC=B)

Given a Reactor Vessel and Internals operating mode and various plant conditions, PREDICT how key system/ plant parameters will respond to the following failures:

a. DBA LOCA

b. Abnormally high water level

c. Abnormally low water level

d. Failed jet pump

e. Failed recirculation pump

f. Shroud access hole cover failure 216000.A2.14 (CFR 41.5 / 45.6)

Ability to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Recirculation flow: Design-Specific (RO=2.9 / SRO=2.9)

Page: 35 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 93 ID: QDC.ILT.16411 Points: 1.00 Unit 1 has a steam leak in the Drywell. Drywell sprays were initiated and secured when required.

Containment parameters are presently:

• Drywell temperature 120°F

• Drywell pressure 1.80 psig and rising

• Torus temperature 91°F

• Torus pressure 1.80 psig and rising The NSO reports the following annunciators are in alarm:

• Annunciator 901-3 C-13, "TORUS VACUUM BRK VALVES OPEN DIV I".

• Annunciator 901-3 G-11, "TORUS VACUUM BRK VALVES OPEN DIV II".

What is the impact, if any, to the primary containment and why?

The primary containment...

A. CAN perform its intended safety function. There is NO impact on primary containment operability.

B. may NOT perform its intended safety function because initial conditions are NOT met for ensuring the maximum drywell pressure during a LOCA will remain below the design value.

C. may NOT perform its intended safety function because initial conditions are NOT met for ensuring the negative differential pressure across the drywell wall will remain below the design value.

D. may NOT perform its intended safety function because initial conditions are NOT met for ensuring that an event producing hydrogen and oxygen does NOT result in a combustible mixture inside the primary containment.

Answer: B Answer Explanation:

Technical Specification bases states that all drywell-to-torus vacuum breakers must be closed to satisfy the pressure-suppression function of the containment.

With annunciators 901-3 C-13(G-11) in alarm, one of the 12 Drywell-Torus vacuum breakers is open and compromising the pressure-suppression function of containment.

Candidate must verify that the alarms for the vacuum breakers are consistent with plant conditions (d/p is zero between the drywell and torus when it should be no less than 0.5 psid).

Distractor 1: Plausible to consider primary containment operable because not all 12 vacuum breakers have to open as intended. Also plausible if candidate assumes that there is a second vacuum breaker in the line (similar to torus-to-reactor building vacuum breakers).

Distractor 2: Plausible because the assumptions for closed vacuum breakers control the amount of negative d/p across the containment walls.

Page: 36 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Distractor 3: Plausible if candidate confuses the alarm indication for the torus-to-reactor building vacuum breakers (i.e. D-14, Torus Vacuum Relief LVL 20B Not Closed)

Reference:

TS B 3.6.1.8 Rev 40, QCAN 901(2)-3 C-13 Rev 9, QCAN 901(2)-3 Rev 7 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 2 Group: 2 Question Source: Modified from Quad Cities ILT Exam Bank (QDC.ILT.15447)

Question History: N/A 10 CFR Part 55 Content: 43(b)(2)

SRO Justification: Candidate must have knowledge of Technical Specification bases and primary containment operability requirements.

Comments: None Associated objective(s):

S-1601-K33 (Freq: LIC=I)

DISCUSS the bases for Containment Systems Tech Spec LCO's.

223001.2.4.46 Ability to verify that the alarms are consistent with the plant conditions.

(RO=4.2 / SRO=4.2)

Page: 37 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 94 ID: QDC.ILT.16412 Points: 1.00 Given:

• Unit 1 is in a refueling outage with fuel offloading scheduled to begin this shift.

• SRM 21 is INOPERABLE and bypassed with its joystick.

You are facing the 901-5 panel full-core display.

Which of the following is true concerning the administrative requirements for fuel moves?

A. Fuel can be offloaded from the LOWER RIGHT quadrant ONLY.

B. Fuel can be offloaded from all quadrants EXCEPT the UPPER LEFT.

C. Fuel can be offloaded from ALL quadrants.

D. Fuel can NOT be offloaded from ANY quadrant.

Answer: B Page: 38 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Answer Explanation:

QCFHP 0100-01, Step E.5.d, states that while you must have 2 operable SRMs, they must be in the quadrant you are performing the core alterations in and in a quadrant adjacent to where core alterations are being performed.

Distractor 1: Plausible if candidate assumes that two SRMs are required in an adjacent quadrant.

Distractor 2: Plausible if the candidate assumes that an SRM in an adjacent quadrant meets the requirements.

Distractor 3: Plausible if candidate assumes that SRMs are required in all fuel regions during an offload.

Reference:

QCFHP 0100-01 Rev 31 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 3 Group: N/A Question Source: Quad Cities ILT Exam Bank (QDC.ILT.15608)

Question History: N/A 10 CFR Part 55 Content: 43(b)(6)

SRO Justification: Candidate must have knowledge of administrative requirements associated with moving fuel.

Comments: None Associated objective(s):

SRL-805-K19 (Freq: LIC=I NF=I)

Given Refueling related equipment operability or key parameter indications and various plant conditions, DETERMINE, from memory, if the Conduct of Refueling Tech Spec LCOs have been met.

2.1.40 Knowledge of refueling administrative requirements. (RO=2.8/ SRO=3.9)

Page: 39 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 95 ID: QDC.ILT.16488 Points: 1.00 Refer to the ACTIONS Table provided in Example 1.3-2 to answer the following question.

Assume that the system is a two (2) pump system.

• Pump A is declared inoperable at 0800 on November 7.

• Pump B is declared inoperable at 1330 on November 7.

Which one of the following is required to maintain compliance with Technical Specifications?

A. Immediately exit Condition A and immediately enter LCO 3.0.3.

B. Immediately enter LCO 3.0.3 and continue to track Condition A completion time from 0800 on November 7.

C. Re-enter Condition A at 1330 on November 7 for Pump B and track its time separately. LCO 3.0.3 entry is NOT required.

D. Restore both pumps to OPERABLE status by 0800 on November 14. LCO 3.0.3 entry is NOT required.

Answer: B Answer Explanation:

Operation with more than 1 pump inoperable is not defined in the LCO. Therefore, LCO 3.0.3 must be entered while continuing to track Condition A time.

Distractor 1: Plausible if candidate assumes that once LCO 3.0.3 is entered, completion time of Condition A stops.

Distractor 2: Plausible if candidate assumes that seperate entry condition is allowed.

Incorrect because separate entry condition is not stated as being allowed.

Distractor 3: Plausible if candidate assumes that required actions of Condition A is adequate for any number of inoperable pumps.

Reference:

TS Completion Times Section 1.3 Amendment 199/195 Reference provided during examination: TS Section 1.3 Example 1.3-2 Cognitive level: High Level (RO/SRO): SRO Tier: 3 Group: N/A Question Source: Quad Cities Bank (QDC.L.102648)

Question History: N/A 10 CFR Part 55 Content: 43(b)(2)

SRO Justification: Candidate must apply Required Actions in accordance with rules of application requirements of Section 1.

Comments: None Page: 40 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Associated objective(s):

Given the Improved Technical Specifications (ITS) and the associated Bases, the trainee shall:

GIVEN PLANT CONDITIONS, APPLY THE RULES OF ITS SECTION 1.3 TO ENSURE COMPLIANCE WITH TECHNICAL SPECIFICATIONS.

2.2.40 Ability to apply Technical Specifications for a system. (RO=3.4 / SRO-4.7)

Page: 41 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 96 ID: QDC.ILT.16416 Points: 1.00 Unit 2 was at 100% power when an accident occurred.

• QGA 200, Primary Containment Control, is being executed.

Current plant conditions require VENTING the primary containment to stay below the Primary Containment Pressure Limit (PCPL).

• QGAs allow exceeding release rate limits

• Drywell pressure is 52 psig and rising

• Torus pressure is 50 psig and rising

• Torus water level is 16 feet Which of the following is the PREFERRED method of venting the drywell given the current plant conditions?

A. Vent the TORUS via the hardened vent ONLY long enough to maintain containment pressure below the PCPL limit.

B. Vent the TORUS via the hardened vent and MAINTAIN a continuous vent path until containment pressure is restored to its normal range.

C. Vent the DRYWELL via the hardened vent ONLY long enough to maintain containment pressure below the PCPL limit.

D. Vent the DRYWELL via the hardened vent and MAINTAIN a continuous vent path until containment pressure is restored to its normal range.

Answer: A Answer Explanation:

The preferred method of venting the primary containment is via the torus. A decontamination factor of up to four is provided when vented gases are required to bubble through water first. As long as torus water level is less than 30 ft, the torus can be vented.

A precaution in QCOP 1600-03, Post-Accident Venting of the Primary Containment, states that there should be an attempt to limit the total amount of the radioactive release by controlling and maintaining Containment pressure below its applicable limit, rather than maintaining a continuous vent path. This precaution does NOT show up in the procedure flow chart or hard card that the RO would use to vent the containment. It is the Unit Supervisors responsibility to set the pressure band for venting using the hardened vent.

Distractor 1: Plausible if candidate assumes that once the limit is reached, you must maintain a continuous vent condition.

Distractor 2: Plausible because drywell pressure will be higher than torus pressure and venting the drywell will give a more immediate pressure reduction than by venting through the torus. Also plausible if candidate assumes that the high torus water level condition prevents venting the torus.

Distractor 3: Combination of distractor 1 and 2.

Page: 42 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Reference:

QCOA 1600-13 Rev 22 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 3 Group: N/A Question Source: Monticello ILT Exam Bank Question History: Monticello 2007 ILT NRC Exam 10 CFR Part 55 Content: 43(b)(4)

SRO Justification: Candidate must have knowledge of the radiation hazards that apply during hardened vent operation (emergency situation). Also supported by a SRO-only facility knowledge objective.

Comments: None Associated objective(s):

S-0001-K24 (Freq: LIC=B)

Given QGA 200, 'Primary Containment Control' and QGA 200-5, 'Hydrogen Control', and various conditions, EVALUATE the conditions and DESCRIBE how to proceed through the flowcharts including transitions within QGA 200 or 200-5, to other QGA procedures or to normal operating procedures.

2.3.11 Ability to control radiation releases. (RO=3.8 / SRO=4.3)

Page: 43 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 97 ID: QDC.ILT.16406 Points: 1.00 Given the following:

• At 12:00 the threshold value for an UNUSUAL EVENT was exceeded, with indication available in the control room.

• At 12:08 the Shift Emergency Director classified an UNUSUAL EVENT based on the associated control room indication.

Based on the above conditions, which of the following identifies the LATEST time at which the State/Local notifications of the UNUSUAL EVENT classification can be initiated and still meet notification requirements per EP-AA-114 "Notifications"?

A. 12:15 B. 12:23 C. 13:00 D. 13:08 Answer: B Answer Explanation:

Candidate must know that the Nuclear Accident Reporting System (NARS) is a telecommunication network and form used to transmit information to appropriate state and local agencies. This notification must be initiated within 15 minutes of the declaration of an emergency.

Distractor 1: Plausible because the event must be classified within 15 of the threshold value being exceeded.

Distractor 2: Combination of distractor 1 and 2.

Distractor 3: Plausible if candidate confuses when the NRC must be notified (ENS) and when state and local agencies must be notified.

Reference:

EP-AA-114 Rev 8 Reference provided during examination: None Cognitive level: High Level (RO/SRO): SRO Tier: 3 Group: N/A Question Source: Modified from LaSalle Exam Bank Question History: LaSalle 2008 ILT NRC Exam 10 CFR Part 55 Content: Unique to the SRO position SRO Justification: Candidate must have knowledge of the performance requirements of the Shift Emergency Director (SED).

Comments: None Page: 44 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Associated objective(s):

2.4.29 Knowledge of the emergency plan. (RO=3.1 / SRO=4.4)

EP Training (G-5) Objective 7 Given event specific information, determine when the state and local agencies must be notified.

Page: 45 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 98 ID: QDC.ILT.16409 Points: 1.00 Complete the following two statements regarding interpretation and execution of Quad Cities Emergency Operating Procedures (QGAs):

Per the QGA Marking Standards, an ARROW pointing at a step in a QGA leg indicates that the crew __(1)__.

Per OP-QC-103-102-1002, Quad Cities Strategies for Successful Transient Mitigation, when executing a leg of the QGAs, all steps should be __(2)__.

A. (1) has COMPLETED the referenced step (2) FOLLOWED in order, even if an emergency depressurization parameter is EXCEEDED further down in the leg B. (1) has COMPLETED the referenced step (2) OMITTED up to the blowdown step if an emergency depressurization parameter is EXCEEDED further down in the leg C. (1) is MAINTAINING or WAITING for a specific plant condition (2) FOLLOWED in order, even if an emergency depressurization parameter is EXCEEDED further down in the leg D. (1) is MAINTAINING or WAITING for a specific plant condition (2) OMITTED up to the blowdown step if an emergency depressurization parameter is EXCEEDED further down in the leg Answer: C Answer Explanation:

When actions have progressed to a point where the crew is maintaining or waiting for a specific plant condition, an arrow should be used to mark their place.

When executing a leg of the QGAs, then all steps should be followed in order to allow the use of all of the mitigating systems to be used and their effectiveness assessed, even if a blowdown parameter is exceeded further down in the leg.

Distractor 1: Plausible because there is a marking standard to indicate a step is complete (line through step).

Distractor 2: Combination of distractor 1 and 2.

Distractor 3: Plausible if candidate assumes that a blowdown must be executed even if the blowdown parameter is exceeded upon entry into the EOP.

Reference:

QCAP 0200-10 Rev 41, OP-QC-103-102-1002 Rev 4 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): SRO Tier: 3 Group: N/A Question Source: New Question History: N/A Page: 46 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 10 CFR Part 55 Content: Unique to the SRO position SRO Justification: Execution standards of the EOPs are a function of the SRO (see facility objective).

Comments: None Associated objective(s):

S-0001-K03 (Freq: LIC=I)

DESCRIBE the QGA flowchart procedure structure 2.1.20 Ability to interpret and execute procedure steps. (RO=4.6 / SRO=4.6)

Page: 47 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 99 ID: QDC.ILT.16408 Points: 1.00 Operations needs to place one valve for the Turbine Lube Oil system in a DIFFERENT position than its normal valve lineup to isolate a small leak.

• The leak is NOT threatening to the continued operation of the Unit.

• This task will NOT have an affect on UFSAR described design functions and NO emergency or equipment damage will result.

• NO approved procedure exists for the evolution.

In ADDITION to an Equipment Status Tag (EST), which of the following (if any) is required to support the immediate re-positioning of the valve?

A. No documentation is required.

B. Degraded Equipment Log (DEL)

C. Adverse Condition Monitoring Plan (ACMP)

D. Abnormal Component Positioning Sheet (ACPS)

Answer: D Answer Explanation:

For situations, excluding routine operation, where a component, system or structure is required to be placed in a position differing from its normal line-up, the alignment must be done utilizing an Abnormal Component Positioning Sheet (ACPS). The ACPS will document proper evaluation, performance and restoration of the alignment, ensuring plant configuration control is maintained.

Such situations will occur when it is desired to reposition a component and NO approved documentation exists.

Distractor 1: Plausible because some activities do NOT fall under the controls of OP-AA-108-101 and therefore do not require a procedure. These activities (e.g. swapping MSIV Room Cooler strainers) are listed in OP-QC-108-1001, Activities That Do Not Require Procedural Guidance. This activity is not one listed in OP-QC-108-1001.

Distractor 2: Plausible because Degraded Equipment Logs are used to track degraded and inoperable Systems, Structures and Components (SSCs). Incorrect because this activity does not fall within the requirements of OP-AA-108-104, Technical Specification Compliance, for degraded equipment (i.e. not tied to Tech Specs, ATR, TRM, ISFSI or ODCM requirements).

Distractor 3: Plausible because ACMPs address plant conditions and parameters that are abnormal. Incorrect because they address significant plant conditions that have not yet reached plant operating procedure action levels. ACMPs are not used to control the abnormal positioning of components for configuration control.

Reference:

OP-AA-108-101 Rev 8 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): SRO Page: 48 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Tier: 3 Group: N/A Question Source: Quad Cities ILT Bank (QDC.LWQ.147386)

Question History: N/A 10 CFR Part 55 Content: 43(b)(3)

SRO Justification: Candidate must have knowledge of the processes for changing the plant configuration and the method of documentation.

Comments: None Associated objective(s):

2.2.14 Knowledge of the process for controlling equipment configuration or status.

(RO=3.9 / SRO=4.3)

SRNLF-CM-K4 (Freq: LIC=I NF=I)

Given OP-AA-108-101, Operations Configuration Control, DESCRIBE the process and form used to place a component, system, or structure into a position differing from its normal alignment.

Page: 49 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion) 100 ID: QDC.ILT.16407 Points: 1.00 Which of the following, without prior NRC approval, is a violation of the Conditions and Limitations in the Facility License?

A. Operation in Mode 1 with a Reactor Coolant System unidentified leakage rate of 3 gpm.

B. Control Room staff of one Senior Reactor Operator and two Reactor Operators with Unit 1 in Mode 4 and Unit 2 in Mode 5.

C. Steady state thermal power below the license thermal power limit, with momentary indications above the limit as a result of normal process fluctuations.

D. A change is made to an Appendix R procedure that results in 80 hours9.259259e-4 days <br />0.0222 hours <br />1.322751e-4 weeks <br />3.044e-5 months <br /> to achieve Cold Shutdown.

Answer: D Answer Explanation:

The 10 CFR 50 Appendix R requirement is to achieve cold shutdown in 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. The Facility license allows changes to the fire protection program without NRC approval if those "changes WOULD NOT adversely affect the ability to achieve and maintain safe shutdown in the event of a fire."

Distractor 1: The Technical Specification limit for unidentified RCS leakage is < 5 gpm.

(Reference TS LCO 3.4.4 (b) ). Operation is within the TS LCO and NOT a violation of the Facility Operating License.

Distractor 2: Lineup is within guidelines of Tech Spec 5.2.2 and 10 CFR 50.54(m)(2)( i ).

Distractor 3: Power fluctuations as a result of normal oscillations due to stabilizing after control rod movement or flow adjustments are NOT considered "intentional operation above the Facility License limit of 2957 MWth".

Reference:

10 CFR50 Appendix R, DPR 29 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): SRO Tier: 3 Group: N/A Question Source: Quad Cities ILT Bank (QDC.ILT.15556)

Question History: Quad Cities 2009 ILT NRC Exam 10 CFR Part 55 Content: 43(b)(1)

SRO Justification: Candidate must have knowledge of the administration of the fire protection program requirements.

Comments: None Page: 50 of 51 29 March 2011

EXAMINATION ANSWER KEY Quad Cities 2011 ILT NRC Exam (SRO Portion)

Associated objective(s):

2.4.25 Knowledge of fire protection procedures. (RO=3.3 / SRO=3.7)

S-ARP-K027 (Freq: LIC=B)

ANALYZE a given condition that may impact the operability of Safe Shutdown Systems or System Combinations (ie component/controller failure, Clearance) using P&ID/C&IDs, E-prints, QCARPs and QCAP 1500-02, if necessary, and DETERMINE if the Safe Shutdown Systems or System Combinations are operable.

Page: 51 of 51 29 March 2011