ML12083A186

2012 Quad Cities Nuclear Power Station Initial License Examination Proposed RO Written Exam

ML12083A186
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Site:	Quad Cities
Issue date:	05/04/2012
From:	NRC/RGN-III/DRS/OLB
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EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 1 of 154 29 February 2012 1

ID: QDC.ILT.15669 Points: 1.00 Unit 2 was at full power when the 2A Recirc pump tripped.

NO operator actions are taken.

The following indications are observed:

96 10 9

8 7

6 5

4 3

2 1

0 71 71 71 71 71 67 63 61 70 100 50 60 70 80 90 100 Time after Recirc Pump Trip (sec)

Reactor Power (%)

Given the above indications, the LARGEST contributor to the drop in reactor power is due to...

A.

core void fraction INCREASING.

B.

DEHC stabilizing reactor pressure.

C.

'D' Feedwater Heater outlet temperature lowering.

D.

Total Steam Flow decreasing.

Answer:

A Answer Explanation:

As total core flow decreases, the coolant's ability to cool the core decreases. However, total steam flow remains constant. When void fraction increases, neutron moderation decreases. This results in more resonance capture and increase in neutron leakage from the core. Void coefficient adds negative reactivity, and reactor power decreases.

Distractor 1 is incorrect: Plausible because a decrease in reactor pressure will add negative reactivity and cause reactor power to drop, however DEHC is acting to raise reactor pressure, not lower it.

Distractor 2 is incorrect: Plausible because the drop in reactor power will cause Feedwater temperature to lower, however the question is asking what is cause of the power drop, not it's effects.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 2 of 154 29 February 2012 Distractor 3 is incorrect: Plausible because eventually total steam flow will lower due to the drop in reactor power as the steam cycle responds to the transient, however, during the first few seconds after the recirc pump trips, total steam flow is constant.

Reference:

BWR Generic Fundamentals - Thermodynamics / Chapter 8 / Thermal Hydraulics Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295001.AK3.02 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION Reactor power response (RO=3.7 / SRO=3.8) 10 CFR Part 55 Content: 41.5 Question Source: New Question History: N/A Comments:

Associated objective(s):

SR-0202-K22 (Freq: LIC=B) Given a Reactor Recirculation System operating mode and various plant conditions, PREDICT how key Reactor Recirculation / plant parameters (including power/flow map shifts) will respond to the following failures:

a. RRCS major failure

b. Reactor recirc pump seal failure (one or both)

c. Reactor recirc pump trip (one or both)

d. Single loop operation with operating pump speed above/below 40%

e. Jet pump or shroud access hole cover failure

f. Recirc flow controller fails high or low

g. Recirc AC oil pump trip

h. Recirc vent fan trip

i. Reactivity additions

j. Core instabilities exist

i. Reactivity additions

j. Core instabilities exist 295001.AK3.02 Reactor power response (RO=3.7 / SRO=3.8)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 3 of 154 29 February 2012 2

ID: QDC.ILT.17154 Points: 1.00 Unit 2 was operating at 100% power when the U2 Reserve Aux Transformer (T-22) experienced an electrical fault resulting in a Phase-C open circuit.

The subsequent drop in voltage on T-22 caused many loads to trip on undervoltage, leading to an automatic reactor scram.

The operators then tripped T-22, resulting in a complete Loss of Offsite Power.

Both the Unit 2 and 1/2 EDGs automatically started and loaded to their respective busses.

Assuming NO other operator action, which of the following is correct for the given conditions?

A.

Turbine Control Valves maintain RPV pressure B.

Feedwater pumps maintain RPV water level C.

RBCCW pumps maintain cooling flow to RBCCW loads D.

TBCCW pumps maintain cooling flow to TBCCW loads Answer:

C Answer Explanation:

A loss of offsite power is a complete loss of all AC busses. When the operators open the feeder breakers to the RAT, the Emergency Diesel Generators will automatically start and power the safety related busses. Only the RBCCW pumps will automatically restart and maintain RBCCW flow.

Question based on recent OPEX from Byron station (1/30/2012) in which an unevaluated transformer fault resulted in a Loss of Offsite Power and Automatic reactor scram. Quad Cities uses the same protective relay structure for the Reserve Aux Transformer as Byron does. Reference IR#01320006, "U2 TRIP SAT C-PHASE FOUND OPEN".

Distractor 1 is incorrect: Plausible if candidate does not recognize the EHC pumps are not automatically re-energized upon restoration of the safety related busses.

Distractor 2 is incorrect: Plausible because the condensate pumps automatically restart upon restoration of the safety related busses.

Distractor 3 is incorrect: Plausible because the TBCCW pumps will automatically restart upon a restoration of power to their supply busses, however, the TBCCW pumps are powered from non-safety related busses that do not automatically have power restored by the EDGs.

Reference:

QCOA 6100-03 rev 28, Byron OPEX IR#01320006, "U2 TRIP SAT C-PHASE FOUND OPEN" Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295003.AA2.04 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER :

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 4 of 154 29 February 2012 System lineups (RO=3.5 / SRO=3.7) 10 CFR Part 55 Content: 41.10 Question Source: Modeled after a question on the 2010 Oyster Creek ILT NRC Exam Question History: N/A Comments:

Associated objective(s):

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 295003.AA2.04 System lineups (RO=3.5 / SRO=3.7)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 5 of 154 29 February 2012 3

ID: QDC.ILT.15678 Points: 1.00 Unit 2 is operating at 100% Reactor power when power is lost to 125 VDC Reactor Building Distribution Panel 2.

What action (if any) is required to restore control power to 4KV Bus 23-1 from a safety related power supply?

A.

An Operator must unlock and reposition a keylock switch, then close the reserve feed breaker.

B.

An Operator must transfer the control power fuse block to the reserve holder, then close the reserve feed breaker from Turbine Building 125VDC Bus 2B-1 C.

An Operator must transfer the control power fuse block to the reserve holder, then close the reserve feed breaker from Turbine Building 125VDC Bus 1B-1 D.

NO action is necessary; control power to Bus 23-1 automatically transfers to the alternate supply via an Automatic Transfer Switch.

Answer:

B Answer Explanation:

One control power fuse block is provided at 4KV Bus 23-1. It is removed from the normal control power supply cubicle (from RB Dist Panel 2) and transferred to Cubicle 9. Then the feed at 125 VDC Bus 2B-1 is closed.

Distractor 1 is incorrect: Plausible because Kirk Key switches are used in other DC applications where there can be no crossties between Div 1 and Div 2.

Distractor 2 is incorrect: Plausible because Turbine Building Reserve Bus 1B-1 is the backup control power supply to 4KV Bus 13-1.

Distractor 3 is incorrect: Plausible because control power will automatically transfer to SBO Bus 7A-1. However, this is not a safety related bus.

Reference:

QOA 6900-13 Loss of Power to Reactor Building Distribution Panel 2, Rev.

15. LN-6900, DC Distribution and Batteries, Rev. 16 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 KA: 295004 Partial or complete loss of DC Power (RO= 4.3 / SRO=4.4) 2.1.23 Ability to perform specific system and integrated plant procedures during all modes of operation 10 CFR Part 55 Content: 41.10 Question Source: New Question History: N/A Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 6 of 154 29 February 2012 Associated objective(s):

295004.2.1.23 Ability to perform specific system and integrated plant procedures during different modes of plant operation. (RO=3.9 / SRO=4.0)

SRN-6900-K26 (Freq: LIC=B NF=B)

EVALUATE given key Station DC Electrical 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. Battery charger trip

b. DC breaker trip

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 7 of 154 29 February 2012 4

ID: QDC.ILT.15671 Points: 1.00 Unit 2 was at 26% Reactor power when the Main Turbine Tripped.

Which one of the following describes how the Reactor responds to this event?

A.

The Reactor scrams as a direct result of the Main Turbine trip.

B.

The Reactor scrams due to high RPV pressure caused by the Main Turbine trip.

C.

Reactor power lowers due to the increase in feedwater temperature.

D.

Reactor power rises due to a decrease in feedwater temperature.

Answer:

D Answer Explanation:

When the Turbine trips, Reactor power is low enough that the direct automatic scram is bypassed. Steam flow is also well within the capacity of the Turbine Bypass valves, so Reactor pressure is affected very little.

The loss of the Turbine does cause a loss of all feedwater heating, causing lower water temperature to be injected to the reactor, inserting positive reactivity and causing reactor power to rise.

Distractor 1 is incorrect: An automatic Reactor scram on a Turbine trip is bypassed when Reactor power is < 38.5% rated thermal power.

Distractor 2 is incorrect: The Turbine trip does not cause a large enough pressure transient to cause RPV pressure to rise to the reactor scram setpoint at this reactor power. This answer would be correct if the reactor failed to scram on the direct Turbine trip at a higher power.

Distractor 3 is incorrect: The Turbine trip causes a loss of extraction steam to the feedwater heaters, which will result in less heat input into the feedwater.

Reference:

QOA 900-5 H-4 Rev 9, QOA 5600-4 rev 27 Reference provided during examination: N/A Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295005.AA2.05 Ability to determine and/or interpret the following as they apply to MAIN TURBINE GENERATOR TRIP :

Reactor power. (RO=3.8 / SRO=3.9) 10 CFR Part 55 Content: 41.10 Question Source: Modified from 2009 Columbia Generating Station ILT NRC Exam Question History: N/A Comments: None.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 8 of 154 29 February 2012 Associated objective(s):

SR-5600-K24 (Freq: LIC=B)

Given a Main Turbine and Auxiliary Systems operating mode and various plant conditions, PREDICT how each supported system will be impacted by the following Main Turbine and Auxiliary Systems failures:

a. Turbine trip

b. Loss of lube oil pressure 295005.AA2.05 Reactor power. (RO=3.8 / SRO=3.9)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 9 of 154 29 February 2012 5

ID: QDC.ILT.15672 Points: 1.00 Unit 2 was operating at 100% power with the 2B FRV in manual when the 2B ASD tripped, causing RPV water level to rise.

At 44 inches Reactor water level, the Unit Supervisor orders a manual reactor scram.

The operators insert a manual scram and place the Reactor Mode Switch in 'SHUTDOWN'.

The highest Reactor water level reached was 44 inches.

Assuming NO other operator action, which one of the following describes how reactor pressure is controlled after the reactor scram?

(Consider each item separately)

A.

Turbine Control Valves throttle to control reactor pressure until the Turbine trips, and then the Turbine Bypass Valves throttle to control pressure.

B.

The Turbine Bypass Valves immediately throttle to control reactor pressure C.

ADS Valves open, then close and remain closed Turbine Bypass Valves throttle to control reactor pressure D.

ADS Valves cycle to control reactor pressure Answer:

A Answer Explanation:

RPV level does not reach the Turbine trip setpoint, and therefore the Turbine remains online. DEHC will attempt to maintain reactor pressure at it's current Pressure Set value (normally 920). With the reactor scrammed, the loss of heat and steam generation combined with the continued injection of feedwater will cause reactor pressure to drop.

The Turbine Control valves will throttle closed to maintain reactor pressure.

As reactor pressure drops, there is less steam being admitted to the Turbine, and the Main Generator will eventually trip on reverse power, immediately tripping the turbine.

Decay heat will raise reactor pressure, which will then be controlled by the Bypass valves.

The safety relief valves will not open as reactor pressure drops rapidly when the scram is inserted.

Decay heat will cause RPV pressure to rise, and the BPVs will reopen sequentially to control reactor pressure.

This question is based on a recent Quad Cities event on Unit 2. Reference IR 01102590.

Distractor 1: Incorrect but plausible if the candidate assumes the turbine trips directly on the reactor scram.

Distractor 2: Incorrect but plausible if the candidate assumes that reactor pressure would rise high enough to open the ADS Valves.

Distractor 3: Incorrect but plausible. This would be the correct answer if the MSIVs closed.

Reference:

Quad Cities specific OPEX reference IR 01102590, QCGP 2-3 rev 74 Reference provided during examination: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 10 of 154 29 February 2012 Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295006.AK2.07 Knowledge of the interrelations between SCRAM and the following:

Reactor pressure control (RO=4.0 / SRO=4.1) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A Comments:

Associated objective(s):

SR-0500-K21 (Freq: LIC=B)

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

a. Panel 901(2)-15/17 (1) RPS trip channel test keylock switches (2) RPS breakers

b. Panel 901(2)-16 (1) Individual scram switches

c. Panel 901(2)-5 (1) Reactor Mode Switch (2) Manual scram pushbuttons (3) Scram Reset Switch (4) Discharge Volume High Water Level Bypass Switch (5) Discharge Volume Isolation test Switch

d. RPS Power Distribution (1) MG control switch (2) Voltmeter Transfer Switch (3) Voltage Adjust rheostat (4) Auxiliary Reset pushbutton (5) RPS supply breakers to 901(2)-15/17 (6) RPS Normal/Reserve power breakers 295006.AK2.07 Reactor pressure control (RO=4.0 / SRO=4.1)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 11 of 154 29 February 2012 6

ID: QDC.ILT.15674 Points: 1.00 The Control Room has been abandoned.

Unit 2 Reactor Water level indication is available at...

A.

The 2202-29 rack in the HPCI room.

B.

The 1/2 2251-104 panel in the Safe Shutdown Makeup Pump room.

C.

The Nematron display panel in the Feedwater Regulating Valve area.

D.

The 2202-5 and 2202-6 racks on the second floor of the reactor building.

Answer:

D Answer Explanation:

Reactor level instrumentation is available at the 5 & 6 instrument racks per QOA 0010-05 rev 24. None of the other areas have water level indication, but are injection control stations.

Distractors 1 through 3 are incorrect: Plausible because they list RPV water injection stations.

Reference:

QOA 0010-05 rev 24 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295016.AA1.06 Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT :

Reactor water level (RO=4.0/ SRO=4.1) 10 CFR Part 55 Content: 41.7 Question Source: Quad Cities ILT Bank (Question ID 12892)

Question History: N/A Comments:

Associated objective(s):

SRN-EVAC-K09 (Freq: LIC=B NF=B)

Given a Control Room Evacuation, STATE the locations to which your job position may be assigned.

295016.AA1.06 Reactor water level (RO=4.0 / SRO=4.1)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 12 of 154 29 February 2012

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 13 of 154 29 February 2012 7

ID: QDC.ILT.15679 Points: 1.00 Both Units are at full power when Annunciator 912-1 E-1, RX BUILDING COOLING WATER HIGH TEMP, is received on Unit 1.

Operators then CLOSE the MO 1-3701, U-1 RBCCW HDR ISOL.

Which of the following sets of loads on Unit 1 is now isolated that would NOT have been isolated on Unit 2 if the MO 2-3701, U-2 RBCCW HDR ISOL, were CLOSED?

A.

Radwaste Floor Drain Filter Holding Pump Cooler; Radwaste Waste Collector Filter Holding Pump Cooler B.

Reactor Building Equipment Drain Tank Heat Exchanger; Fuel Pool Cooling Heat Exchangers C.

Reactor Water Cleanup System Non-Regenerative Heat Exchangers; Reactor Water Cleanup Pump Seal and Bearing Coolers D.

Drywell Pneumatic Compressor Heat Exchanger; Containment Pumpback System (Joy) Air Compressors and Aftercoolers Answer:

A Answer Explanation:

Per QCOA 3700-03, closing Unit 1 or Unit 2 RBCCW Header Isolation Valves, MO-1(2)-

3701 could become necessary during a RBCCW High temperature transient.

The Radwaste Floor Drain Filter Holding Pump Cooler and Radwaste Waste Collector Filter Holding Pump Cooler are Unit 1 loads only.

Distractors 1, 2 & 3 are incorrect: These are loads isolated by both Units isolation valves. Selected if candidate recognizes them as unique to only Unit 1 but associates them with the incorrect unit.

Reference:

QCOA 3700-03 rev 6 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group:1 K/A: 295018.2.2.03 (multi-unit) Knowledge of the design, procedural, and operational differences between units. (RO=3.1 / SRO=3.3) 10 CFR Part 55 Content: 41.5 Question Source: New Question History: N/A Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 14 of 154 29 February 2012 Associated objective(s):

295018.2.2.03 (multi-unit) Knowledge of the design, procedural, and operational differences between units. (RO=3.1 / SRO=3.3)

SR-3700-K21 (Freq: LIC=B)

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

a. MO 3701 control switch

b. MO 3702, 3703, 3706 control switch

c. Pump control switches

d. DIV I/II DW CLR / RBCCW / FPC TRIP BYPASS switch

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 15 of 154 29 February 2012 8

ID: QDC.ILT.17072 Points: 1.00 Unit 1 is at rated power with the Instrument Air Compressors (IAC) in the following lineup:

Based on the above indications, what is the FIRST response as Instrument Air pressure drops towards 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.

The 1/2 IAC will start in an attempt to restore Instrument Air pressure.

B.

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

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.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 16 of 154 29 February 2012 Answer:

A Answer Explanation:

The picture shows the Instrument Air Compressors in their normal operating lineup, with both the 1/2 and 1/2B IAC control switches in N-A-C, and the 1A and U2 IAC control switches in P-T-L. The 1/2B 'ON' light is lit, indicating that it is currently running. The 1/2

'OFF' light is lit, indicating it is in automatic pressure control mode and will start and load when pressure at discharge of the compressor reaches 95 psig.

The setpoint for 912-1 D-11 is 85 psig. Therefore, the first action to occur would be for the 1/2 IAC to start and load in an attempt to restore Instrument Air pressure.

Distractor 1 is incorrect: 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 is incorrect: Plausible because 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, and if candidate assumes that the 1/2 IAC is tripped or otherwise will not automatically start.

Distractor 3 is incorrect: 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 K/A: 295019.AA1.01 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR :

Backup air supply (RO=3.5 / SRO=3.3) 10 CFR Part 55 Content: 41.7 Question Source: Modified from Quad Cities 2011 ILT NRC Exam Question History: N/A Comments: None Associated objective(s):

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 295019.AA1.01 Backup air supply (RO=3.5 / SRO=3.3)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 17 of 154 29 February 2012 9

ID: QDC.ILT.15677 Points: 1.00 Unit 1 is offline. Shutdown Cooling is in progress using the "A" loop of RHR. Current conditions are as follows:

RPV water level is 30 inches and steady Both Reactor Recirc pumps are running Main Condenser backpressure indicates 24 in Hg (6 in Hg vacuum)

Reactor Recirc pump suction temperature is 265ºF An electric fault causes the MO 1-1001-47, SDC HDR Downstream SV, to CLOSE. All attempts to re-open the valve have failed. RPV temperature starts to rise.

Under these conditions, which of the following methods should be used to cooldown the RPV, and what is the reason for using that method?

A.

Increase RWCU reject in order to increase heat removal from the vessel B.

Open the Turbine Bypass Valves since this is the preferred method for rejecting heat from the reactor C.

Increase RWCU reject to prevent thermal stratification of the RPV D.

Establish Torus Cooling, close the MSIVs and open a relief valve because a total loss of Shutdown Cooling has occurred Answer:

A Answer Explanation:

When a loss of shutdown cooling occurs, other methods for decay heat removal must be utilized to prevent RPV stratification and repressurization. QCOA 1000-02, Loss of Shutdown Cooling, lists the different ways to remove heat from the vessel.

The conditions given in the stem indicate that the condenser is unavailable, along with the normal flowpath from RHR.

With the Recirc pumps still on, the chance for stratification is minimized, however, the need for heat removal still exists.

Distractor 1 is incorrect: Plausible because this would be the correct answer if the condenser was available.

Distractor 2 is incorrect: Plausible because this would be the correct answer if the Recirc pumps were off.

Distractor 3 is incorrect: Plausible because Alternate Shutdown Cooling is entered if there is a loss of Shutdown Cooling occurs AND no other means of removing heat from the RPV are available.

Reference:

QCOA 1000-02 rev 17 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 18 of 154 29 February 2012 K/A: 295021.AK3.04 Knowledge of the reasons for the following responses as they apply to LOSS OF SHUTDOWN COOLING :

Maximizing reactor water cleanup flow (RO=3.3 / SRO=3.4) 10 CFR Part 55 Content: 41.5 Question Source: Modified from LaSalle 2011 ILT NRC Exam Question History: N/A Comments: None Associated objective(s):

SR-1000-K22 (Freq: LIC=B)

Given an RHR system operating mode and various plant conditions, PREDICT how key RHR/RHRSW system and/or plant parameters will respond to the following RHR/RHRSW system component or controller failures:

a. RHR pump trip

b. RHRSW pump trip

c. RHR heat exchanger biofouling

d. RHR minimum flow valve fails open or closed 295021.AK3.04 Maximizing reactor water cleanup flow (RO=3.3 / SRO=3.4)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 19 of 154 29 February 2012 10 ID: QDC.ILT.17136 Points: 1.00 A core reload is in progress on Unit 1 and a fuel assembly is being lowered into the core near SRM 24.

Per QCFHP 0110-02, INADVERTENT CRITICALITY DURING FUEL MOVES, which of the following indications meets the definition of a criticality event in progress during the fuel move?

(Consider each indication SEPARATELY)

1. Rise and subsequent lowering of SRM 24 period as the bundle is being lowered.

2. Sustained increase in SRM 24 count rate after the bundle is set.

3. Multiple spikes in SRM 24 count rate after the bundle is set.

A.

2 ONLY B.

1 and 2 ONLY C.

2 and 3 ONLY D.

1, 2 and 3 Answer:

A Answer Explanation:

Answer: True criticality is indicated by a sustained increase in count rate, over 15 to 20 seconds, of the SRM closest to the Fuel Assembly/Bundle.

Distractor 1 is incorrect: Plausible if candidate assumes that counts rising (not sustained) is an indication of inadvertent criticality.

Distractor 2 is incorrect: Plausible if candidate does not realize that this is the SRM response when masked by power supply spiking, welding, etc.

Distractor 3 is incorrect: Combination of distractor 1 and 2.

Reference:

QCFHP 0110-02 Rev 4 Reference provided during examination: N/A Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295023 Refueling Accidents 2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (RO=3.4 / SRO=3.7) 10 CFR Part 55 Content: 41.7 Question Source: Quad Cities 2009 ILT NRC Exam Question History: Quad Cities 2009 ILT NRC Exam Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 20 of 154 29 February 2012 Associated objective(s):

SRLF-805-K15 (Freq: LIC=B NF=B)

Given symptoms and indications depicting an abnormal/emergency condition during refueling operations, ANALYZE the indications and LOCATE the appropriate abnormal/emergency procedure.

295023.2.4.47 Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material. (RO=3.4 / SRO=3.7)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 21 of 154 29 February 2012 11 ID: QDC.ILT.17115 Points: 1.00 Given the following annunciators:

1. 901-5 D-11, PRIMARY CNMT HIGH PRESSURE

2. 901-3 A-16, PRI CNMT HIGH PRESSURE

3. 901-3 G-4, DRYWELL HIGH PRESSURE

4. 912-7 A-1, DRYWELL 1 N2 LINE PRESS HI Which of the above alarms, when initially received, indicates that QGA 200, Primary Containment Control, is required to be entered?

(Consider each alarm SEPARATELY)

A.

1 and 4 ONLY B.

1 and 3 ONLY C.

2 and 3 ONLY D.

2 and 4 ONLY Answer:

B Answer Explanation:

The initiation setpoints for the alarms are as follows:

(Analytical values) 901-5 D-11, PRIMARY CNMT HIGH PRESSURE 2.5 psig 901-3 A-16, PRI CNMT HIGH PRESSURE 1.5 psig 901-3 G-4, DRYWELL HIGH PRESSURE 2.5 psig 912-7 A-1, DRYWELL 1 N2 LINE PRESS HI 1.5 psig Candidates often see all four annunciators alarming simultaneously at 2.5 psig Drywell pressure. Though there are significantly different plant responses at 1.5 psig and 2.5 psig drywell pressure.

Distractor 1 is incorrect: Plausible because 912-7 A-1 will often be in alarm when 901-5 D-11 is in alarm. However, the question is asking the initiation setpoint of the alarm, not which alarms will be in at 2.5 psig.

Distractor 2 is incorrect: Plausible because 901-3 A-16 will be in alarm when 901-3 G-4 is in alarm. However, the question is asking the initiation setpoint of the alarm, not which alarms will be in at 2.5 psig.

Distractor 3 is incorrect: Plausible because the listed alarms will often be in at 2.5 psig DW pressure. However, the question is asking the initiation setpoint of the alarm, not which alarms will be in at 2.5 psig.

Reference:

QCAN 901-5 D-11 rev 11, 901-3 A-16 rev 12, 901-3 G-4 rev 8, 912-7 A-1 rev 3

Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 22 of 154 29 February 2012 K/A: 295024 High Drywell Pressure 2.4.31 Knowledge of annunciators alarms, indications, or response procedures. (RO 4.2

/ SRO 4.1) 10 CFR Part 55 Content: 41.10 Question Source: New Question History: N/A Comments:

Associated objective(s):

295024.2.4.31 Knowledge of annunciators alarms, indications, or response procedures.

(RO 4.2 / SRO 4.1) (CFR: 41.10)

SR-1601-K06 (Freq: LIC=I)

Given a Containment Systems 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.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 23 of 154 29 February 2012 12 ID: QDC.ILT.17075 Points: 1.00 Unit 2 is at rated conditions when the Turbine Control Valve #1 drifts CLOSED.

In response to the event, Reactor power will initially (1), followed by (2).

A.

(1) RISE (2) the remaining Turbine Control Valves and/or Turbine Bypass Valves going OPEN B.

(1) RISE (2) only Turbine Control Valve #2 going OPEN C.

(1) LOWER (2) the remaining Turbine Control Valves and/or Turbine Bypass Valves going OPEN D.

(1) LOWER (2) only Turbine Control Valve #2 going OPEN Answer:

A Answer Explanation:

The rise in RPV pressure will cause voids to collapse, adding positive reactivity due to the void coefficient, which will then cause reactor power to increase, further raising RPV pressure. The remaining Turbine Control Valves (TCVs) and Turbine Bypass Valves will respond to the pressure rise and throttle open to control RPV pressure.

This question is based on a recent event at Quad Cities on Unit 2. Reference IR 01216208.

Distractor 1 is incorrect: All of the remaining TCVs will open in response to the event and the BPVs will open if necessary to control RPV pressure. Response of the TCVs is not limited to only TCV #2.

Distractor 2 is incorrect: Power will RISE due to the increase in RPV pressure from closure of TCV #1.

Distractor 3 is incorrect: Power will RISE due to the increase in RPV pressure from closure of TCV #1.

Reference:

QOA 5650-02 rev 10, Quad Cites specific OPEX, Reference IR 01216208.

Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295025.EK1.01 Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR PRESSURE :

Pressure effects on reactor power. (RO=3.9 / SRO=4.0) 10 CFR Part 55 Content: 41.8 Question Source: LaSalle ILT Exam Bank, Quad Cites specific OPEX Reference IR 01216208.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 24 of 154 29 February 2012 Question History: LaSalle 07-01 NRC ILT Exam Comments:

Associated objective(s):

SR-5652a-K15 (Freq: LIC=I) DESCRIBE the operation of the following principal Main Turbine Control - EHC Logic System components:

a. Pressure control unit (1) Pressure regulators (2) Pressure/load gate (3) Load limit gate

b. Bypass control unit (1) Bypass jack (2) Max combined flow limit

c. Load control unit (1) Load set (2) Load set runback

d. Speed and acceleration control unit (1) Speed control section (2) Acceleration control section

e. Valve flow control unit

f. Main turbine valves during turbine startup/runback/overspeed (1) Main stop valves (2) Control valves (3) Combined intermediate valves (4) Bypass valves (5) Extraction non-return valve 295025.EK1.01 Pressure effects on reactor power. (RO=3.9 / SRO=4.0)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 25 of 154 29 February 2012 13 ID: QDC.ILT.15644 Points: 1.00 Compared to its normal heat capacity, the ability of the Torus Suppression Pool to condense steam during a DBA LOCA is reduced by which of the following initial conditions?

A.

Torus water level of 15 ft.

B.

Torus water temperature of 102°F C.

Drywell pressure of 2.7 psig D.

Stuck closed Rx Bldg-to-Torus Vacuum Breaker Answer:

B Answer Explanation:

Torus temperature of 102°F is above normal (must maintain < 95°F normally). If initial Torus temperature is high, the Torus suppression capability (condensing steam) is reduced because the energy content of the Torus is higher.

Distractor 1 is incorrect: High Torus water level could result in excessive clearing loads and pool swells during a DBA LOCA.

Distractor 2 is incorrect: A high initial drywell pressure will result in a higher drywell pressure during a DBA LOCA.

Distractor 3 is incorrect: Plausible because this would be the correct answer if the vacuum breaker was stuck open.

Reference:

TS LCO 3.6.2.1 Amendment 199/195 Reference provided during examination: N/A Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295026.EK1.02 Knowledge of the operational implications of the following concepts as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE :

Steam condensation. (RO=3.5 / SRO=3.8) 10 CFR Part 55 Content: 41.8 Question Source: Quad ILT Bank (QDC.ILT.15507)

Question History: Quad Cities 2009 ILT NRC Exam Comments: None.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 26 of 154 29 February 2012 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 295026.EK1.02 Steam condensation. (RO=3.5 / SRO=3.8)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 27 of 154 29 February 2012 14 ID: QDC.ILT.15642 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 158ºF and rising slowly Based on 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 re-started 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 bypassed on the 912-5 panel using keylock switches.

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

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

Distractor 3 is incorrect: 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 K/A: 295028.EA1.03 Ability to operate and/or monitor the following as they apply to HIGH DRYWELL TEMPERATURE :

Drywell cooling system (RO=3.9 / SRO=3.9) 10 CFR Part 55 Content: 41.7 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.16442)

Question History: Quad Cities 2011 ILT NRC Exam Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 28 of 154 29 February 2012 Associated objective(s):

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 295028.EA1.03 Drywell cooling system (RO=3.9 / SRO=3.9)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 29 of 154 29 February 2012 15 ID: QDC.ILT.17121 Points: 1.00 What is the concern for operating RCIC below 11 feet Torus water level?

A.

RCIC turbine speed oscillations and unstable operation due to excessive cycling of the RCIC Turbine Exhaust Check Valve.

B.

Damage to the RCIC turbine exhaust line due to excessive cycling of the RCIC Turbine Exhaust Check Valve.

C.

RCIC High Exhaust Pressure (and eventual turbine trip) due to the RCIC Turbine Exhaust discharging directly to the Torus airspace.

D.

Overpressurization and subsequent failure of the primary containment due to the RCIC Turbine Exhaust discharging directly to the Torus airspace.

Answer:

C Answer Explanation:

The Torus condenses the RCIC turbine exhaust steam. If the torus were to lose its condensing ability due to low level, the turbine exhaust pressure would increase until a turbine trip would occur on high exhaust pressure.

Distractor 1 is incorrect: Plausible because this is a precaution in QCOP 1300-02, RCIC Manual Startup, but the reason for the caution is based on operating RCIC below 2200 rpm.

Distractor 2 is incorrect: Plausible because this is a precaution in QCOP 1300-02, RCIC Manual Startup, but the reason for the caution is based on operating RCIC below 400 gpm.

Distractor 3 is incorrect: Plausible because this is the reason HPCI is secured before reaching a torus level of 11 ft.

Reference:

QCOP 1300-02 rev 30, LN-1300 rev 10 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295030 EK3.03 Knowledge of the reasons for the following responses as they apply to LOW SUPPRESSION POOL WATER LEVEL:

RCIC operation: Plant-Specific (RO=3.6 / SRO=3.7) 10 CFR Part 55 Content: 41.8 Question Source: New Question History: N/A Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 30 of 154 29 February 2012 Associated objective(s):

SR-1300-P02 (Freq: LIC=A)

Given a reactor plant in an accident condition with drywell pressure above 2.5 psig, start RCIC for pressure control in accordance with QCOP 1300-02.

295030.EK3.03 RCIC operation: Plant-Specific (RO=3.6 / SRO=3.7)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 31 of 154 29 February 2012 16 ID: QDC.ILT.15683 Points: 1.00 Given:

A valid automatic Reactor scram from full rated power has occurred All but four control rods are fully inserted to at least position 04 Drywell pressure is 9 psig and rising slowly RPV pressure is being manually controlled between 800 to 1000 psig using ADS valves RPV water level is -142 inches and lowering slowly A Group IV Isolation has occurred Based on the above conditions, which of the listed systems have automatically started and are currently injecting to the RPV?

1. RCIC

2. HPCI

3. Core Spray A.

1 only B.

2 only C.

3 only D.

1, 2 and 3 Answer:

A Answer Explanation:

HPCI and RCIC auto start and inject automatically when RPV water level falls to -59 inches. However, HPCI has isolated due to the Group IV isolation. Core Spray will have auto started, but the cannot inject until RPV pressure is less than 325 psig.

Distractor 1 is incorrect: Plausible if candidate incorrectly identifies a Group IV isolation as a RCIC isolation Distractor 2 is incorrect: Plausible because Core Spray will have started but doesn't inject until RPV pressure is less than 325 psig.

Distractor 3 is incorrect: Plausible if candidate incorrectly identifies a Group IV isolation.

Reference:

QCAN 901(2)-4 D-16 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295031.EK2.04 Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the following:

Reactor core isolation cooling: Plant-Specific (RO=4.0 / SRO=4.1) 10 CFR Part 55 Content: 41.7

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 32 of 154 29 February 2012 Question Source: New Question History: N/A Comments:

Associated objective(s):

295031.EK2.04 Reactor core isolation cooling: Plant-Specific (RO=4.0 / SRO=4.1)

SR-1300-K07 (Freq: LIC=B)

LIST the signals which cause a RCIC System auto initiation including setpoints.

DESCRIBE how they are reset.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 33 of 154 29 February 2012 17 ID: QDC.ILT.17057 Points: 1.00 In accordance with QGA 101, during an ATWS, Boron injection must be initiated before Torus water temperature reaches.

A.

95°F B.

105°F C.

110°F D.

120°F Answer:

C Answer Explanation:

The combination of high reactor power, high torus temperature and high drywell pressure, are symptomatic of heat being rejected to the torus at a rate in excess of that which can be removed by the torus cooling system.

Unless mitigated, these conditions ultimately result in loss of NPSH for ECCS pumps taking suction on the torus, containment overpressurization, and ultimately to the loss of primary containment integrity.

If torus temperature cannot be held below the Heat Capacity Temperature Limit (HCTL),

QGA 200 will require a blowdown. It is desirable to shut down the reactor before the blowdown is required. Boron is therefore injected when the torus heatup begins. If power is relatively low, boron injection can be completed and the reactor shut down before torus temperature reaches the HCTL.

Distractor 1 is incorrect: Plausible because 95°F is a QGA 200 entry condition.

Distractor 2 is incorrect: Plausible because 105°F is associated with Torus water temperature Tech Spec limit 3.6.2.1 Condition C.

Distractor 3 is incorrect: Plausible because 120°F is associated with Torus water temperature Tech Spec limit 3.6.2.1 Condition E.

Reference:

Lesson plan L-QGA101 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295037 EA2.04 Ability to determine and/or interpret the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN :

Suppression pool temperature. (RO=4.0 / SRO=4.1) 10 CFR Part 55 Content: 41.10 Question Source: Modified from Vermont Yankee 2009 NRC exam Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 34 of 154 29 February 2012 Associated objective(s):

SR-0001-K58 (Freq: LIC=B)

Define the following QGA related terms:

a. Hot Shutdown Boron Weight

b. Cold Shutdown Boron Weight

c. Boron Injection Initiation Temperature 295037.EA2.04 Suppression pool temperature. (RO=4.0 / SRO=4.1)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 35 of 154 29 February 2012 18 ID: QDC.ILT.15643 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 is incorrect: 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 is incorrect: 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 is incorrect: 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 K/A: 295038.EK1.02 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) 10 CFR Part 55 Content: 41(8-10)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 36 of 154 29 February 2012 Question Source: Quad Cities ILT Exam Bank (QDC.ILT.15618)

Question History: Quad Cities 2011 ILT NRC Exam Comments: None Associated objective(s):

295038.EK1.02 Protection of the general public (RO=4.2 / SRO=4.4)

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

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

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 37 of 154 29 February 2012 19 ID: QDC.ILT.15686 Points: 1.00 Unit 1 and Unit 2 are operating at 100% power with the following indications at the 912-5 panel:

RED status light is illuminated for CR A HVAC AHU GREEN status light is illuminated for CR B HVAC AHU A FIRE occurs in the Service Building and detectors sense smoke in the RETURN air duct to the A Train AHU.

With NO operator action, which of the following identifies an expected indication at the 912-5 panel?

RED status light is illuminated for...

A.

CR A HVAC AHU ONLY B.

CR B HVAC AHU ONLY C.

CONTROL ROOM HVAC AFU 1/2A D.

CONTROL ROOM HVAC AFU 1/2B Answer:

A Answer Explanation:

Answer: With smoke detected in an Emergency zone (via the return air duct), the A Train AHU will automatically shift into 100% Smoke Purge Mode (No recirculation) Therefore, the A Train continues to run and the B Train remains in standby.

Distractor 1 is incorrect: Plausible if the candidate believes that the Emergency Train

("B" Train) is aligned when smoke is detected.

Distractor 2 and 3 are incorrect: The AFU supplies filtered outside air to whichever AHU is running, however, there are no auto-starts for the AFUs.

Reference:

QCOP 5750-09 Rev 48 Reference provided during examination: N/A Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 600000.AA1.05 Ability to operate and / or monitor the following as they apply to PLANT FIRE ON SITE:

Plant and control room ventilation systems (RO=3.0 / SRO=3.1) 10 CFR Part 55 Content: 41.7 Question Source: Grand Gulf ILT Exam Bank Question History: 2008 GG ILT NRC Exam Comments: None.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 38 of 154 29 February 2012 Associated objective(s):

600000.AA1.05 Plant and control room ventilation systems (RO=3.0 / SRO=3.1)

SR-5752-K20 (Freq: LIC=B)

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

a. A train (1) Compressor suction/ discharge pressures (2) Compressor oil level and pressure (3) Chill water temperatures (4) Hot water temperatures (5) Area static pressures and temperatures (6) Run status lights:

(a) Compressors (b) Chill water pumps (c) Convertor pumps

b. B train AFU (1) Combined, pre/post HEPA, prefilter differential pressures (2) Temperatures and flow (3) Booster fan run status (4) Cooling water flow

c. B train AHU fan run status

d. RCU compressor run status

e. Control room differential pressure, humidity, temperatures

f. Damper positions

g. Freon gas leak detection power/trouble LEDs

h. Toxic gas analyzer NH3 concentration

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 39 of 154 29 February 2012 20 ID: QDC.ILT.17074 Points: 1.00 Unit 2 was operating at full power when the following occurs:

Generator Differential Overcurrent Generator Negative Phase Sequence Which one of the following describes the response of the plant?

A.

The Main Turbine AND Generator Field Breaker TRIP The Exciter Field Breaker and Isophase Bus Duct Cooling Fans are UNAFFECTED B.

The Main Turbine, Generator Field Breaker AND Isophase Bus Duct Cooling Fans TRIP The Exciter Field Breaker is UNAFFECTED C.

Main Turbine, Generator Field Breaker, Exciter Field Breaker AND Isophase Bus Duct Cooling Fans TRIP D.

Main Turbine, Generator Field Breaker AND Exciter Field Breaker TRIP Isophase Bus Duct Cooling Fans are UNAFFECTED Answer:

D Answer Explanation:

Generator Negative Phase Sequence and Generator Differential Overcurrent cause a Main Generator Lockout Relay to actuate, which will trip the Turbine, the Generator Output GCBs, and the Generator Field Breaker (which then trips the Exciter Field Breaker).

Distractor 1 is incorrect: Plausible because the Exciter Field Breaker is not directly tripped by the Lockout Relay. The Exciter Field Breaker is interlocked to trip when the Generator Field Breaker trips.

Distractor 2 is incorrect: Combination of distractors 1 and 3.

Distractor 3 is incorrect: Plausible because the Isophase Bus Duct Cooling Fans are required to be on when the Generator Output Breakers are closed. However, there is not direct trip due to the Generator.

Reference:

QCAN 901(2)-8 A-12 rev 2, LN-6000 rev 10 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 700000.AK2.02 Knowledge of the interrelations between GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES and the following:

Breakers, relays (RO=3.1 / SRO=3.3) 10 CFR Part 55 Content: 41.7 Question Source: Hope Creek ILT Exam Bank

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 40 of 154 29 February 2012 Question History: Hope Creek 2009 ILT NRC Exam Comments:

Associated objective(s):

700000.AK2.02 Breakers, relays (RO=3.1 / SRO=3.3)

SR-6000-K10 (Freq: LIC=I)

DESCRIBE how the Main Generator System responds to the following trips:

a. Generator lockout (86)

b. Generator backup lockout (86B)

c. Output circuit breakers

d. Generator field breaker

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 41 of 154 29 February 2012 21 ID: QDC.ILT.17061 Points: 1.00 Unit 1 is at 100% power Annunciator 901-54 C-7 NORMAL PROCESS FLOW HI/LO alarms.

Investigation shows the following indication:

Which of the following transients is most likely to cause this indication?

A.

Loss of condenser vacuum caused by the SJAE suctions closing.

B.

Loss of condenser vacuum caused by a blown SJAE 35 foot loop seal.

C.

Recombination occurring in the off-gas piping upstream of the air ejectors' flow element.

D.

An explosion within the off-gas system downstream of the recombiner.

Answer:

B Answer Explanation:

Picture shows a steady increase in off-gas flow, caused by an increase in airflow from the blown loop seal.

Distractor 1 is incorrect: Closure of the SJAE suctions would cause flow to lower.

Distractor 2 is incorrect: Recombination occurring upstream of the flow element at the air ejectors would result in no change at the holdup pipe or beyond.

Distractor 3 is incorrect: An explosion in the off-gas system would cause erratic off-gas flow indication. The picture shows a consistent increase in flow.

Reference:

QOA 3300-02 rev 38, LOSS OF CONDENSER VACUUM Reference provided during examination: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 42 of 154 29 February 2012 Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 295002.AA2.04 Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM :

Offgas system flow. (RO=2.8 / SRO=2.9) 10 CFR Part 55 Content: 41.10 Question Source: New Question History: N/A Comments:

Associated objective(s):

295002.AA2.04 Offgas system flow. (RO=2.8 / SRO=2.9)

SR-5400-K22 (Freq: LIC=B)

Given an Off-Gas System operating mode and various plant conditions, PREDICT how key Off-Gas System/ plant parameters will respond to the following Off-Gas System component or controller failures:

a. Loss of recombination

b. Recombination at a location other than the recombiner

c. Off-gas fire or explosion

d. SJAE pressure regulator failure

e. Preheater steam supply pressure regulator failure

f. Off-gas charcoal adsorber fire

g. Improperly operated SJAE manual suction and steam supply valves

h. Blown loop seals (1) SJAE intercondenser drain 35 foot loop seal (2) 30 min holdup line drain loop seal (3) Final filter drain loop seal

i. Mechanical vacuum pump plugged Cuno filter

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 43 of 154 29 February 2012 22 ID: QDC.ILT.17152 Points: 1.00 Unit 1 was in normal full power operation.

Five (5) Unit 1 CRDMs were operating at greater than 350°F due to leaking scram outlet valves.

Which of the following incomplete-scram conditions could result?

A.

Scram Discharge Volume hydraulic lock B.

Loss of rod position indication C.

Failure to latch at position 00 D.

Slow rod scram insertion times Answer:

D Answer Explanation:

Elevated temperatures above 350°F can slow the scram insertion time of the affected CRD due to water in the drive over piston area flashing to steam during a scram when the scram outlet valve opens and vents the over piston area to the Scram Discharge Volume. Rapid expansion of the steam/water mixture tends to increase backpressure and slow the start of rod motion.

Distractor 1 is incorrect: Plausible because leakage past the scram outlet valves allow water to flow to the SDV, however the SDV is continuously drained to the RBEDT.

Distractor 2 is incorrect: Plausible if candidate assumes the CRD limit switches will fail if they get too hot.

Distractor 3 is incorrect: Plausible because stop piston damage can occur due to excessive CRDM drive pressures.

Reference:

LIC 0301 rev 10, QCOS 0300-21 rev 10 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 2 K/A: 295015.AK2.07 Knowledge of the interrelations between INCOMPLETE SCRAM and the following:

CRD mechanism. (RO=3.3 / SRO=3.4) 10 CFR Part 55 Content: 41.7 Question Source: Modified from DAEC 2005 ILT NRC Exam Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 44 of 154 29 February 2012 Associated objective(s):

295015.AK2.07 CRD mechanism. (RO=3.3 / SRO=3.4)

SR-0302-K22 (Freq: LIC=B)

Given a Control Rod Drive Hydraulics operating mode and various plant conditions, PREDICT how key system/plant parameters will respond to the following Control Rod Drive Hydraulics component or controller failures:

a. CRD pump trip

b. CRD FCV valve failure open/close

c. Leaking scram valves

d. Excessive accumulator gas pressures

e. Excessive charging water pressure

f. Overpiston flowpath isolated

g. Leaking directional control valves (insert or withdrawal)

h. Low cooling water flow

i. High cooling water pressure

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 45 of 154 29 February 2012 23 ID: QDC.ILT.17099 Points: 1.00 (Refer to the Reactor Pressure/Level/Steam Flow traces provided)

The plant was operating at 2957 MWth when a transient occurred.

Based on the RPV pressure, water level, and steam flow traces provided, SELECT the transient that CAUSED the observed response.

(Assume NO operator action unless otherwise stated)

A.

Stuck open ERV, Manual Reactor scram B.

Turbine trip, Automatic Reactor scram C.

All RFPs trip, Automatic Reactor Scram D.

Group I isolation, Automatic Reactor scram Answer:

D Answer Explanation:

Three Graphs:

RPV pressure curve shows initial drop of approx. 25 psig due to the anticipatory scram on 10% MSIV closure, followed by a pressure spike to approx. 1030 psig.

Pressure then lowers to 910 psig due to decay heat. Curve stops at 1007 psig and still slowly rising.

Total Steam Flow curve shows a drop to zero and no return due to the MSIV closure.

RPV water level curve shows an initial drop to -10 inches (void collapse followed by a gradual return to normal level.

Distractor 1 is incorrect: A reactor manual scram with a stuck open ERV would actually show only a slight increase in reactor pressure followed by a small pressure drop. RPV pressure then would continually lower slowly due to the open ERV.

Distractor 2 is incorrect: Turbine trip with an automatic reactor scram would show a larger initial pressure spike, ie. > 1070 psig. The other major difference is the RPV pressure will be controlled at 920 psig steam dome as can be seen at t = 4 min.

Distractor 3 is incorrect: If all RFPs trip, reactor water level will quickly drop to the scram setpoint and the reactor will scram. However, there will be no reactor pressure spike associated with the scram.

Reference:

QCAN 901(2)-5 B-4 rev 7 Reference provided during examination: Set of 3 graphs reflecting Reactor water level, pressure and steam flow traces.

Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 2 K/A: 295020.AK3.04 Knowledge of the reasons for the following responses as they apply to INADVERTENT CONTAINMENT ISOLATION:

Reactor pressure response. (RO=4.1 / SRO=4.1)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 46 of 154 29 February 2012 10 CFR Part 55 Content: 41.5 Question Source: Quad Cities ILT Exam Bank Question History: Quad Cities 2009 ILT Cert Exam Comments:

Associated objective(s):

SR-0250-K22 (Freq: LIC=B)

Given a Main Steam System operating mode and various plant conditions, PREDICT how system/plant parameters will respond to the following Main Steam System failures:

a. MSL leak inside the drywell

b. MSL leak outside the drywell

c. MSIV closure (one or more lines) 295020.AK3.04 Reactor pressure response. (RO=4.1 / SRO=4.1)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 47 of 154 29 February 2012 24 ID: QDC.ILT.17063 Points: 1.00 Refer to the indications on the following page.

Units 1 and 2 are operating at full power.

A moment ago, Unit 1 received alarm 901-3 E-1, RX BLDG VENT CHANNEL A DOWNSCALE.

Before Operators took any action with regard to annunciator 901-3 E-1, in quick succession Unit 2 received the following alarms:

902-3 A-3 RX BLDG VENT CH A OR B HI RADIATION 902-3 H-3 RX BLDG VENT CHANNEL B HI HI RADIATION Based on the above conditions, which of the following indications depicted would you expect to see for Unit 1 and Unit 2 Reactor Building ventilation?

A.

1 B.

2 C.

3 D.

4 Answer:

D Answer Explanation:

A Hi-Hi radiation alarm on either unit's reactor building ventilation radiation monitor causes the isolation of both units' reactor building ventilation systems.

Distractors 1, 2 and 3 are incorrect: All of the distractors show incorrect damper positions.

Reference:

QCAN 901(2)-3 H-3 rev 9, QCOP 5750-02 rev 21 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 2 K/A: 295034.EA1. Ability to operate and/or monitor the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION :

295034.EA1.03 Secondary containment ventilation. (RO=4.0 / SRO=3.9) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 48 of 154 29 February 2012 Associated objective(s):

SR-5750-K20 (Freq: LIC=B) Given a Plant Ventilation Systems operating mode and various plant conditions, EVALUATE the following Plant Ventilation Sstems indications/responses and DETERMINE if the indication/ response is expected and normal.

a. Reactor building ventilation (1) Differential pressures (2) Damper positions (3) Building supply/exhaust fan status and amperage (4) Supply, exhaust and outside air temperatures

b. Turbine building ventilation (1) Differential pressures (2) Damper positions (3) Building supply/exhaust fan status and amperage (4) Main chimney flow rate (5) East/west supply, and exhaust air temperatures

c. Radwaste building ventilation:

(1) Building supply/exhaust fan and damper status (2) Differential pressures, temperatures

d. Off-gas filter building ventilation (1) Fan status and damper positions (2) Differential pressures, temperatures and flow (3) Digital Mimic Display on Panel 2212-37B(A)

(a) Flashing Red Light (computer hardware failure)

e. Off-gas filter building freon leak detector indication 295034.EA1.03 Secondary containment ventilation. (RO=4.0 / SRO=3.9)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 49 of 154 29 February 2012 25 ID: QDC.ILT.17130 Points: 1.00 Unit 2 is in Mode 5 with the following evolutions in progress:

Under-vessel work that has been categorized as an Operation with the Potential to Drain the Reactor Vessel (OPDRV).

Spent fuel is being moved within the Fuel Pool.

Unit 1 is at rated power with New Fuel being channeled.

A spent Fuel Dry Cask is being lowered to the Reactor Building Trackway.

Annunciator 912-5 C-1, RX BLD 1 LOW DP, has just alarmed.

The ANSO reports:

Differential Pressure Indicator (DPI) 1-5740-22, RX BLDG TO ATMOS DP, reads 0 inches of water Two (2) Reactor Building Supply Fans and two (2) Reactor Building Exhaust Fans are operating on each Unit.

Given the above conditions, which of the following actions are required in order to comply with LCO 3.6.4.1 Secondary Containment?

Immediately initiate action to suspend...

A.

spent fuel Dry Cask transfer in the Trackway.

B.

spent fuel moves in the Unit 2 Fuel Pool.

C.

New Fuel Channeling on Unit 1.

D.

OPDRVs on Unit 2.

Answer:

D Answer Explanation:

OPDRVs can be postulated to cause significant fission product release to the secondary containment. In this case, the Reactor Building is the only barrier to release of fission products to the environment.

With Reactor Building DP at or above 0 inches of water, Secondary Containment integrity is no longer assured. The immediate suspension of OPDRVs is required in order to minimize the probability of a vessel draindown and subsequent potential for fission product release.

Distractor 2 is incorrect: Plausible because this would be correct if the the fuel being moved in the fuel pool were recently irradiated fuel.

Distractors 1 and 3 are incorrect: Plausible because there are many procedural restrictions placed on these evolutions, but no Tech Spec required actions.

Reference:

LCO 3.6.4.1 Secondary Containment amendment no. 245/240 Reference provided during examination: None Cognitive level: Memory

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 50 of 154 29 February 2012 Level (RO/SRO): RO Tier: 1 Group: 2 K/A: 295035.EK1.01 Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE:

Secondary containment integrity (RO=3.9 / SRO=4.2) 10 CFR Part 55 Content: 41.9 Question Source: New Question History: N/A Comments: OTPS challenged the use of a "low d/p" situation in the question as the K/A statement is for "high d/p". Facility Rep and Exam Author acknowledged the comment and reviewed the comment with the NRC Chief Examiner.

The NRC Chief Examiner granted permission to use the question with a "low d/p" situation in the question vs. a "high d/p" situation, as there is no K/A statement for this system that addresses "low d/p". The operational implications for a "low d/p" situation are significant because the secondary containment is kept at a negative pressure with regards to the outside environment.

Associated objective(s):

295035.EK1.01 Secondary containment integrity (RO=3.9 / SRO=4.2)

SR-1601-K29 (Freq: LIC=I)

Given Containment Systems key parameter indications and various plant conditions, DETERMINE, from memory, if the Containment Systems Tech Spec LCOs have been met.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 51 of 154 29 February 2012 26 ID: QDC.ILT.17118 Points: 1.00 Which of the following conditions, if any, require Unit 1 to enter QGA 300, Secondary Containment Control?

Condition 1: 10" of water in the Unit 1 HPCI room due to fire fighting efforts combating a lube oil fire Condition 2: Automatic isolation of the Unit 1 Reactor Building Ventilation system due to High Drywell pressure A.

None B.

1 only C.

2 only D.

1 and 2 Answer:

B Answer Explanation:

Any Table U area that has a water level above 1 inch requires entry into QGA 300, regardless of the source.

The 10 inches of water reported in the HPCI area therefore requires entry into QGA 300.

Distractor 1 is incorrect: Plausible because there is a decision point in QGA 300 concerning the discharge of a primary system into the RB which is a common misconception that it relates to the entry condition.

Distractor 2 is incorrect: Combination of distractors 1 and 3 Distractor 3 is incorrect: Plausible if the candidate assumes that any isolation of the RB vents require entry into QGA 300. RB vent rads must be > 10 mr/hr

Reference:

QGA 300 rev 12 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 2 K/A: 295036 Secondary Containment High Sump / Area Water Level 2.4.04 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. (RO=4.5 /

SRO=4.7) 10 CFR Part 55 Content: 41.10 Question Source: New Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 52 of 154 29 February 2012 Associated objective(s):

SR-0001-K27 (Freq: LIC=B)

STATE the entry conditions to QGA 300, 'Secondary Containment Control'.

295036.2.4.04 Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. (RO=4.0 / SRO=4.3)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 53 of 154 29 February 2012 27 ID: QDC.ILT.17147 Points: 1.00 The 'CAM/ACAD PWR CONT' Control Switch on the 901-56 panel has been moved from 'AUTO' to 'ON' in order to monitor for Hydrogen in the primary containment.

Why is there a requirement to wait 45 minutes before taking a reading?

A.

To ensure the sample lines are clear of moisture B.

To allow the analyzer to warm up to proper operating temperature C.

To obtain a reliable sample indicative of containment conditions D.

To allow the analyzer to cycle through all 4 channel sample lines Answer:

C Answer Explanation:

A note in QCOP 2400-01, CAM Subsystem Operation, states: "Allow approximately 45 minutes for analyzer indication to stabilize in order to obtain reliable information."

It is expected that the Hydrogen and Oxygen readings will oscillate given the volume of the containment space. 45 minutes allows time for readings to stabilize and therefore obtain a more accurate and representative sample.

Distractor 1 is incorrect: Plausible because moisture in the sample lines could make the obtained readings inaccurate. However, the sample lines are heat traced to prevent condensation buildup.

Distractor 2 is incorrect: Plausible because the analyzers will require time to warm up if they weren't kept warm by the installed and normally on heater boxes.

Distractor 3 is incorrect: Plausible because the very similar, but completely separate Oxygen analyzer system cycles through different sample lines.

Reference:

QCOP 2400-01 rev 19 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 1 Group: 2 K/A: 500000.EA1.01 Ability to operate and monitor the following as they apply to HIGH CONTAINMENT HYDROGEN CONTROL:

Primary containment hydrogen instrumentation (RO=3.4 / SRO=3.3) 10 CFR Part 55 Content: 41.7 Question Source: Hope Creek ILT Exam Bank Question History: Hope Creek 2010 ILT NRC Exam Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 54 of 154 29 February 2012 Associated objective(s):

SR-2400-K15 (Freq: LIC=I)

DESCRIBE the operation of the following principle CAM components:

a. CAM H2-O2 gas analyzer

b. CAM sample line heat trace 500000.EA1.01 Primary containment hydrogen instrumentation (RO=3.4 / SRO=3.3)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 55 of 154 29 February 2012 28 ID: QDC.ILT.15687 Points: 1.00 Given the following plant conditions:

Unit 1 was initially at rated power A Loss of Offsite Power (LOOP) and a small-line-break Loss of Coolant Accident (LOCA) have occurred All control rods have fully inserted Reactor pressure is at 800 psig and lowering slowly RCIC is injecting at 400 gpm but HPCI has tripped and is unavailable RPV water level is -49" and lowering at 10" per minute Drywell pressure is 2.0 psig and rising at 0.5 psig per minute The Unit 1 Emergency Diesel Generator has started and loaded to its respective bus The 1/2 Emergency Diesel Generator has started but HAS NOT loaded onto its respective bus Assuming no operator action, how will the Automatic Depressurization System (ADS) SRVs respond?

The ADS SRVs will A.

NOT automatically open.

B.

automatically open in 60 seconds.

C.

automatically open in 170 seconds.

D.

automatically open in 570 seconds.

Answer:

C Answer Explanation:

The Unit 1 EDG running loaded on to its bus indicates that there is power to half of the available ECCS pumps. With Drywell pressure above 2.5 psig, and RPV level < -59 inches, the 110 second ADS timer will start. The requirements for ADS to auto blowdown will be met if a Low pressure ECCS pump running with the discharge pressure > 100 psig, therefore, the relief valves will automatically open.

Distractor 1 is incorrect: Plausible if candidate does not recognize there is power available to an ECCS pump.

Distractor 2 is incorrect: Plausible if candidate does not realize that the 110 second timer must time out before the ADS can open automatically.

Distractor 3 is incorrect: In 60 seconds the RPV level will be -59", also, if it was present without Drywell >2.5 psig, an 8.5 minute (510 seconds) timer starts and times-out.

Reference:

QCAN 901(2)-3 B-13 rev 7 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 56 of 154 29 February 2012 K/A: 203000.K3.03 Knowledge of the effect that a loss or malfunction of the RHR/LPCI:

INJECTION MODE (PLANT SPECIFIC) will have on following:

Automatic depressurization logic (RO=4.2 / SRO=4.3) 10 CFR Part 55 Content: 41.7 Question Source: Modified from 2009 Quad Cities ILT NRC Exam Question History: N/A Comments: None Associated objective(s):

203000.K3.03 Automatic depressurization logic (RO=4.2 / SRO=4.3)

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.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 57 of 154 29 February 2012 29 ID: QDC.ILT.17073 Points: 1.00 Which of the following actions will bypass ALL RHR Pump Suction Valve interlocks?

1. Placing the 'FUEL POOL CLG SUCTION PERMISSIVE' switch in the 'OVERRIDE' position

2. Placing the 'RHR Loop A/B CONTAINMENT CLG PERMISSIVE SWITCH 17' to 'ON' A.

Neither B.

1 only C.

2 only D.

Both 1 AND 2 Answer:

B Answer Explanation:

The RHR suction valves must be opened in order to satisfy the pump start permissive logic. The purpose of the interlock is to prevent starting a RHR pump without enough NPSH which would in turn damage the pump.

The 'FUEL POOL CLG SUCTION PERMISSIVE' switch will override all the RHR pump suction interlocks.

Distractor 1 is incorrect: Plausible if the candidate assumes there is no way to override the suction interlocks.

Distractor 2 is incorrect: Plausible because the S17 switch will override the containment spray valve interlocks on the discharge side of the RHR pumps.

Distractor 3 is incorrect: Plausible because both switches override RHR pumps interlocks.

Reference:

LN-1000 rev 16, CCOP 1000-11 rev 17 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 203000.K4.06 Knowledge of RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following:

Adequate pump net positive suction head (interlock suction valve open): Plant-Specific (RO=3.5 / SRO=3.5) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 58 of 154 29 February 2012 Associated objective(s):

203000.K4.06 Adequate pump net positive suction head (interlock suction valve open):

Plant-Specific (RO=3.5 / SRO=3.5)

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

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 59 of 154 29 February 2012 30 ID: QDC.ILT.17080 Points: 1.00 Unit 1 was in Mode 5 during a planned refueling outage with Shutdown Cooling in service, when a Loss of Offsite Power occurs. Unit 2 is at 100% power and the U1 EDG is Out-Of-Service.

The ANSO has started and loaded the U1 SBO to Bus 14-1.

Which Unit 1 RHR pumps are available for shutdown cooling?

A.

None B.

A and B only C.

C and D only D.

All Answer:

D Answer Explanation:

Question stem states a loss of off-site power with the U-1 EDG OOS and a LOCA. The response to loss of off-site power will be for the 1/2 EDG to load to bus 13-1. 13-1's RHR pumps (A&B) are available for shutdown cooling.

The stem further states the ANSO starts and loads the U1 SBO to 14-1. 14-1's RHR pumps (C&D) are available for shutdown cooling.

With A&B&C&D RHR pumps available, all RHR pumps are available for shutdown cooling.

Distractor 1 is incorrect: Distracter says A&B only so C&D are not included in the distracter.

Distractor 2 is incorrect: Distracter says C&D only so A&B are not included in the distracter.

Distractor 3 is incorrect: All 4 are available.

Reference:

QOM 1-6500-T04 (rev 8) and T06 (rev 7)

Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 205000.K2.01 Shutdown Cooling System (RHR Shutdown Cooling Mode)

Knowledge of electrical power supplies to the following:

Pump motors (RO=3.1 / SRO=3.1) 10 CFR Part 55 Content: 41.7 Question Source: Quad Cities ILT Exam Bank Question History: Quad Cities ILT NRC Exam for class 03-01 Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 60 of 154 29 February 2012 Associated objective(s):

SR-1000-K19 (Freq: LIC=B)

LIST the plant systems which support the RHR system and DESCRIBE the nature of support. (Includes power supplies) 205000.K2.01 Pump motors (RO=3.1 / SRO=3.1)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 61 of 154 29 February 2012 31 ID: QDC.ILT.17084 Points: 1.00 Unit 2 was recently scrammed due to a loss of offsite power.

HPCI was started in pressure control mode and the following conditions established:

Reactor Pressure is 990 psig and lowering slowly HPCI discharge pressure is 1150 psig and steady The Flow Indicating Controller (FIC) is set at 5600 gpm The FIC is in Automatic HPCI turbine speed is 4000 RPM The Unit Supervisor determines a higher cooldown rate is necessary.

Which of the following methods will raise the cooldown rate?

A.

RAISE the HPCI FIC setpoint to 5700 gpm.

B.

LOWER the HPCI FIC setpoint to 5500 gpm.

C.

Throttle MO 2-2301-10 (Test Return Valve) more OPEN.

D.

Throttle MO 2-2301-10 (Test Return Valve) more CLOSED.

Answer:

D Answer Explanation:

HPCI is providing the means of lowering reactor pressure (i.e., doing the cooldown in the Pressure leg of QGA 100). Raising the cooldown rate raises the rate at which reactor pressure is lowered.

HPCI is operating with the governor limiting speed at 4000 rpm. No increase in speed is possible, so the amount of work done by the pump [and thus the turbine] must be increased. Throttling the MO 2-2301-10 more CLOSED increases the amount of flow restriction in the discharge line, requiring more work from the pump.

Distractor 1 is incorrect: Raising the FIC setpoint would modify the MGU (motor gear unit) position in an attempt to control turbine speed, but with the governor controlling speed at 4000 rpm, no increase in speed is possible. The pump flowpath has not changed, so no change in steam consumption by the HPCI turbine occurs.

Distractor 2 is incorrect: Lowering the FIC setpoint would modify the MGU position in an attempt to control turbine speed. Lowering the MGU position will either have no effect if the governor continues to be the lowest speed device or the turbine would slow if the MGU becomes the lowest speed device. Since the pump flowpath has not changed, no change in steam consumption occurs or less steam consumption occurs if the MGU position lowers sufficiently.

Distractor 3 is incorrect: Throttling 2-2301-10 more open lessens the restriction to flow in the line. As a result, the pump has to do less work to obtain the same flowrate. This in turn lowers the amount of steam necessary to drive the pump and reduces the cooldown rate.

Reference:

QCOP 2300-06, Rev 31 Reference provided during examination: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 62 of 154 29 February 2012 Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 206000.K1. Knowledge of the physical connections and/or cause-effect relationships between HIGH PRESSURE COOLANT INJECTION SYSTEM and the following:

206000.K1 03 Reactor pressure: BWR-2,3,4 (RO=3.8 / SRO=3.8) 10 CFR Part 55 Content: 41.8 Question Source: New Question History: None Comments:

Associated objective(s):

206000.K1 03 Reactor pressure: BWR-2,3,4 (RO=3.8 / SRO=3.8)

SR-2300-K21 (Freq: LIC=B)

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

a. HPCI manual initiation pushbutton

b. Remote turbine trip pushbutton and latch

c. HPCI trip reset switch

d. Turbine trip test pushbutton

e. Isolation trip channel reset keylocks

f. Miscellaneous drain valves reset

g. Thrust bearing test pushbuttons

h. Turning gear controls

i. Turning gear reset switch

j. HPCI flow controller (auto/man operation)

k. Motor speed changer

l. Block Motor speed changer

m. Motor gear unit

n. Gland seal condensate pump

o. Gland seal cooling water pump

p. Gland seal leakoff blower

q. Auxiliary oil pump

r. Emergency bearing oil pump

s. MOV/AOV controls (control room)

t. Room cooler controls

u. Local valve control stations

v. Oil heater controls

w. Min flow valve controller

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 63 of 154 29 February 2012 32 ID: QDC.ILT.17101 Points: 1.00 Which one of the following completes the statement below, IAW QGA 100 RPV CONTROL?

Maintaining RPV water level at the jet pump suction with at least one Core Spray pump injecting into the RPV at (1) gpm provides assurance that (2) exists.

A.

(1) 4500 (2) Adequate Core Cooling B.

(1) 4500 (2) the Minimum Steam Cooling RPV Water Level C.

(1) 5050 (2) Adequate Core Cooling D.

(1) 5050 (2) the Minimum Steam Cooling RPV Water Level Answer:

C Answer Explanation:

Spray Cooling is a method of maintaining Adequate Core Cooling (ACC) and is defined as at least one Core Spray loop injecting at or above 5050 gpm (rated is 4500 gpm per pump at 90 psig reactor pressure), and RPV water level at or above the top of the jet pumps (-191 in).

Distractor 1 is incorrect: Plausible because the rated flow for a Core Spray pump is 4500 gpm at 90 psig reactor pressure.

Distractor 2 is incorrect: Combination of distractors 1 and 3.

Distractor 3 is incorrect: The Minimum Steam Cooling RPV Water Level is -166 inches.

The stem states "...with water level AT the jet pump..." which is < -191 inches.

Reference:

L-QGADET rev 8 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 209001.K5.04 Knowledge of the operational implications of the following concepts as they apply to LOW PRESSURE CORE SPRAY SYSTEM :

Heat removal (transfer) mechanisms (RO=2.8 / SRO=2.9) 10 CFR Part 55 Content: 41.5 Question Source: Brunswick ILT Exam Bank Question History: Brunswick 2010 ILT NRC Exam Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 64 of 154 29 February 2012 Associated objective(s):

SR-0001-K09 (Freq: LIC=B)

DESCRIBE the purpose of the following QGA curves/tables:

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 Detail E, Alternate Injection Systems

d. QGA Detail F, Injection Subsystems

e. QGA Detail G, Preferred ATWS Systems

f. QGA Detail H, Alternate ATWS Systems

g. QGA Table J, Minimum Steam Cooling Pressure

h. QGA Figure K, Drywell Spray Initiation Limit

i. QGA Figure L, Pressure Suppression Pressure

j. QGA Figure M, Heat Capacity Limit

k. QGA Detail O, Emergency Depressurization Systems

l. QGA Detail P, RPV Injection Sources

m. QGA Detail Q, Alternate Flooding Systems

n. QGA Table S, Reactor Building Area Temperatures

o. QGA Table T, Reactor Building Area Radiation Levels

p. QGA Table U, Reactor Building Area Water Levels

q. QCAP 0200-10 Attachments S,T,U,V and W, RHR and CS NPSH Curves

r. QCAP 0200-10 Attachment X, HPCI NPSH Curves

s. QCAP 0200-10 Attachment Y, RCIC NPSH Curves

t. QCAP 0200-10 Attachment Z, ECCS Vortex Limit

u. Cold Shutdown Boron

v. Hot Shutdown Boron

w. Maximum Subcritical Banked Withdrawal Position

x. Minimum Number Of SRVs Required For Emergency Depressurization

y. Minimum Number Of ADS Valves For Decay Heat Removal

z. Decay Heat Removal Pressure aa. Minimum Steam Cooling RPV Water Level ab. Minimum Zero-Injection RPV Water Level 209001.K5.04 Heat removal (transfer) mechanisms (RO=2.8 / SRO=2.9)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 65 of 154 29 February 2012 33 ID: QDC.ILT.17090 Points: 1.00 Unit 1 was operating at 100% power when an ATWS occurred.

Reactor pressure is being maintained at 920 psig with the Turbine Bypass Valves SBLC discharge pressure is cycling between 500 and 600 psig The following indications are observed at the 901-5 panel:

Based on the above indications, what is the NEXT required action in order to inject boron into the reactor?

A.

Inject boron using the RWCU system B.

Place the SBLC Pump Selector switch in the 'SYS 1&2' position C.

Place the SBLC Pump Selector switch in the 'SYS 2&1' position D.

Place the SBLC Pump Selector switch in the 'SYS 2' position Answer:

D Answer Explanation:

The indications are the 'A' SBLC pump is running but the relief valve is lifting early, preventing any flow to the reactor. Because an ATWS is occuring only 1 pump operation is allowed. QCOP 1100-02 states that if flow is not occuring, select SYS 2.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 66 of 154 29 February 2012 Distractor 1 is incorrect: The discharge pressure is inconsistent with the discharge path being blocked, so the correct action would be to select the other system in an attempt to use SBLC, not to use RWCU.

Distractor 2 is incorrect: The 'B' squib valve is not expected to fire when selecting "SYS 1". Procedural direction in QCOP 1100-02 is to select the other system, not both systems.

Distractor 3 is incorrect: The relief valve is lifting early. Procedural direction in QCOP 1100-02 is to select the other system, not both systems.

Reference:

QCOP 1100-02, Rev 12 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 211000.A2.04 Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Inadequate system flow (RO=3.1 / SRO=3.4) 10 CFR Part 55 Content: 41.5 Question Source: New Question History: N/A Comments:

Associated objective(s):

211000.A2.04 Inadequate system flow (RO=3.1 / SRO=3.4)

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

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 67 of 154 29 February 2012 34 ID: QDC.ILT.17104 Points: 1.00 Unit 1 is at rated power with both RPS MG Sets in service and RPS reset.

What will the pressure be at the selected points on the Scram Air Header following a loss of both RPS A and B?

(Assume NO operator action)

Point 1 Point 2 Point 3 A.

85 psig 85 psig 85 psig B.

85 psig 85 psig 0 psig C.

85 psig 0 psig 0 psig D.

0 psig 0 psig 0 psig

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 68 of 154 29 February 2012 Answer:

C Answer Explanation:

When both RPS buses deenergize, all of the scram pilot solenoid valves deenergize and vent the scram valves, allowing scram water from the CRD system to fully insert the rods.

This will also complete the logic for the Backup Scram Valves to energize and not only block the air to the scram valves, but also to depressurize the header downstream of the Backup Scram Valves (point 2), providing an alternate means to scram the rods.

Also on a loss of RPS, the Scram Dump Valves deenergize, venting air from the Scram Discharge Volume vent valves, depressurizing the header downstream of the Scram Dump Valves (point 3).

The Scram Air Header upstream of the Backup Scram Valves will remain pressurized at 85 psig (point 1).

Distractor 1 is incorrect: Plausible if the candidate assumes that the Scram Air Header remains pressurized and only the scram valves depressurize.

Distractor 2 is incorrect: Plausible if the candidate assumes that only scram valves and SDV vent & drain valves solenoids deenergize.

Distractor 3 is incorrect: Plausible if the candidate assumes that the entire Scram Air Header depressurized on a loss of RPS or if the ARI valves were energized.

Reference:

LIC 0500 rev 7, QCOS 0500-05 rev 21 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 212000.K1.15 Knowledge of the physical connections and/or cause-effect relationships between REACTOR PROTECTION SYSTEM and the following:

SCRAM air header pressure (RO=3.8 / SRO=3.9) 10 CFR Part 55 Content: 41.6 Question Source: New Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 69 of 154 29 February 2012 Associated objective(s):

SR-0500-K20 (Freq: LIC=B)

Given a Reactor Protection System operating mode and various plant conditions, EVALUATE the following Reactor Protection System indications/responses and DETERMINE if the indication/ response is expected and normal:

a. Reactor Protection (1) Scram air header pressure (2) Scram valve downstream piping temperature (3) Scram air header valves air leakage (4) Channel A/B scram solenoid group lights (5) Scram valve position lights (6) Manual scram pushbutton lights

b. RPS Power Distribution (1) RPS MG motor breaker indicating lights (2) MG/Bus voltage (3) MG set current (4) EPA lights (5) MG set targets (UF,OV) 212000.K1.15 SCRAM air header pressure (RO=3.8 / SRO=3.9)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 70 of 154 29 February 2012

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 71 of 154 29 February 2012 35 ID: QDC.ILT.17123 Points: 1.00 Which of the following is the power supply to Intermediate Range Monitor (IRM) Ch 18 drawer on the 901-36 panel?

A.

MCC 19-1-1 B.

RPS B C.

24/48 VDC Bus 1B D.

Essential Service Bus Answer:

C Answer Explanation:

The power supply to IRM Ch 18 is Div 2 24/48 VDC, which is Bus 1B.

Distractor 1 is incorrect: Plausible because MCC 19-1-1 supplies power to the IRM drive mechanism motors.

Distractor 2 is incorrect: Plausible because RPS B supplies power to the APRM associated with IRM 18.

Distractor 3 is incorrect: Plausible because the ESS Bus supplies power to the IRM recorders.

Reference:

QOA 6900-03 rev 12, LIC 0702 rev 8 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 215003.K2.01 Knowledge of electrical power supplies to the following:

IRM channels/detectors (RO=2.5 / SRO=2.7) 10 CFR Part 55 Content: 41.7 Question Source: Quad Cities ILT Exam Bank Question History: N/A Comments:

Associated objective(s):

215003.K2.01 IRM channels/detectors (RO=2.5 / SRO=2.7)

SR-0702-K19 (Freq: LIC=I)

LIST the plant systems which support Intermediate Range Monitor System and DESCRIBE the nature of support. (Includes power supplies)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 72 of 154 29 February 2012

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 73 of 154 29 February 2012 36 ID: QDC.ILT.17129 Points: 1.00 A startup is in progress on Unit 1.

IRM CH 15 is bypassed Recirc Loop Temperatures are approximately 200ºF The Reactor reached the POAH on IRM Range 7 All IRMs are exhibiting a trend similar to the one shown (Range 7 / fast speed).

Based on the given indications, what action(s) are required next?

A.

Notch out control rods to maintain a reactor heat up ONLY B.

Range up IRMs ONLY C.

Range up IRMs and notch out control rods to maintain a reactor heat up D.

Insert control rods to shutdown the reactor Answer:

C Answer Explanation:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 74 of 154 29 February 2012 The picture shows an IRM trend recorder with increasing power (from 45 to 60) and then turning, indicating the positive reactivity added by withdrawing control rods is being offset by the negative reactivity being added from the negative moderator temperature coefficient as the coolant heats up. This is expected and the Reactor Operator must continue to withdraw control rods to maintain the reactor heat up.

QCGP 1-1, Normal Unit Startup, specifically warns about an inadvertent subcriticality. "An excessive delay at this point in the startup may result in negative reactivity addition due to fuel and moderator heating, leading to subcriticality." Operators are directed to maintain reactor parameters and IRM count rate, and not allow the count rates to lower.

IRMs are to be ranged up at approximately 50/125 of scale by procedure.

Distractor 1 is incorrect: The IRMs are to be ranged up at 50/125s of scale. The IRM shown is reading at approximately 60/125s.

Distractor 2 is incorrect: The IRMs must be ranged up, however control rods are required to be notched out as well in order to maintain heat up.

Distractor 3 is incorrect: This would be the correct answer if reactor power were below the POAH.

Reference:

QCGP 1-1 rev 87, Page 51. QCOP 0700-02 rev 15 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 215003 Intermediate Range Monitor System 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) 10 CFR Part 55 Content: 41.5 Question Source: New Question History: N/A Comments:

Associated objective(s):

SR-0300-P06 (Freq: LIC=I (ILT-MP) Given a reactor plant during a startup, withdraw control rods to achieve criticality and establish a reactor heatup rate while maintaining the heatup rate and vessel temperatures within TS limits in accordance with QCOP 0280-01, QCGP 1-1, QCGP 4-1 and QCOS 0201-02.

215003.2.1.07 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

(RO=3.7 / SRO=4.4)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 75 of 154 29 February 2012 37 ID: QDC.ILT.17137 Points: 1.00 Given:

Unit 2 was at full rated power Both Manual Scram pushbuttons have been depressed The Reactor Mode Switch has been placed in 'SHUTDOWN' All ARI pushbuttons have been armed and depressed Recirc pumps have been run back to minimum speed ALL Scram Group Solenoid lights are LIT NO other operator actions have been taken The SRM Detector Drive indications are as shown:

Based on the above indications, what is the status of the SRM system?

The 'SRM/IRM DETECTOR POSITION display' switch is (1).

The SRM detectors (2) driving into the core.

A.

(1) DE-SELECTED (2) ARE NOT B.

(1) SELECTED (2) ARE C.

(1) SELECTED (2) ARE NOT D.

(1) DE-SELECTED (2) ARE Answer:

A Answer Explanation:

The picture shows the SRM CHANNEL 21-24 SELECT lights illuminated and all 'IN' &

'OUT' position lights dark.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 76 of 154 29 February 2012 These indications are consistent with the SRM/IRM DETECTOR POSITION display switch de-selected. The Reactor Mode switch in Shutdown with all RPS scram solenoid lights LIT indicate that an electric ATWS has occurred and RPS has not processed the scram. Because RPS is still energized, the SRMs will not automatically drive into the core and must be manually inserted.

Distractor 1 is incorrect: Plausible because this would be the correct answer if there were no Electric ATWS. The SRMs will not drive in until RPS is deenergized or they are manually inserted.

Distractor 2 is incorrect: Plausible because the second part of the distractor is correct.

Distractor 3 is incorrect: Plausible because this answer would also be correct if there were no Electric ATWS.

Reference:

QCOP 0700-01 rev 12 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 1 Group: 1 K/A: 215004.K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the SOURCE RANGE MONITOR (SRM) SYSTEM :

RPS (RO=3.2 / SRO=3.3) 10 CFR Part 55 Content: 41.6 Question Source: New Question History: N/A Comments:

Associated objective(s):

215004.K6.01 RPS (RO=3.2 / SRO=3.3)

SR-0701-K21 (Freq: LIC=B)

Given various plant conditions, PREDICT how Source Range Monitor System/plant parameters will respond to manipulation of the following Source Range Monitor System controls:

a. SRM/IRM detector position display on/off

b. SRM/IRM detector selector switches

c. SRM/IRM Drive In / Drive Out pushbuttons

d. Function switch

e. Bypass Joy Stick

f. Inop Inhibit pushbutton

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 77 of 154 29 February 2012 38 ID: QDC.ILT.17092 Points: 1.00 Unit 1 is operating at full power when Flow Converter #1 fails downscale.

Which ONE of the following describes the expected response to this failure?

A.

OPRM #1 swaps its flow signal to its companion OPRM.

B.

APRM #1's indicated reactor power rises.

C.

A half-scram occurs.

D.

The Rod Worth Minimizer inserts withdrawal and insert Rod Blocks.

Answer:

C Answer Explanation:

A downscale failure of flow converter #1 would cause the reference scram setpoint to lower along with a rod out block. As the APRM's fission rate remains unchanged, the indicated power remains constant but the flow converter causes the APRM's flow-biased scram setpoint to lower, resulting in a half-scram.

Distractor 1 is incorrect: OPRM #1's companion OPRM is OPRM #3 which also uses the same flow converter. This swap would occur only if the companion OPRM had a valid flow signal.

Distractor 2 is incorrect: APRM #1 does not use the flow converter for indicated power, only for the APRM flow biased scram setpoint.

Distractor 3 is incorrect: The Rod Block Monitor system inserts a withdraw block but does not cause an insert block. Rod Worth Minimizer used in distracter to give plausibility to both types of blocks.

Reference:

QCAN 901(2)-5 D-6, Rev 5 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 215005.K1.16 Knowledge of the physical connections and/or cause-effect relationships between AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM and the following:

Flow converter/comparator network: Plant-Specific (RO=3.3 / SRO=3.4) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 78 of 154 29 February 2012 Associated objective(s):

215005.K1.16 Flow converter/comparator network: Plant-Specific (RO=3.3 / SRO=3.4)

SR-0703-K22 (Freq: LIC=B)

Given an LPRM/APRM System operating mode and various plant conditions, PREDICT how the LPRM/APRM System and plant parameters will be impacted by the following failures:

a. Loss of RPS power

b. Flow convertor output fails high/low

c. LPRM output fails high/low

d. APRM output fails high/low

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 79 of 154 29 February 2012 39 ID: QDC.ILT.17125 Points: 1.00 Refer to the RCIC mimic on the following page Unit 1 is at rated power with RCIC in a standby lineup.

You are the assist NSO walking down the panels.

Based on the given indications, which of the following annunciators, if any, is consistent with these conditions?

A.

NO annunciators should be in alarm.

B.

901-4 A-14, RCIC SUCTION HIGH PRESSURE C.

901-4 H-15, RCIC GLAND SEAL VAC TANK HIGH LEVEL D.

901-4 F-16, RCIC TURBINE INLET STM DRN HIGH LEVEL Answer:

C Answer Explanation:

The picture shows the Barometric Condenser Condensate Pump running (red light on) and the Condensate Pump Isolation valves, AO 1-1301-12 & 13 open (Red lights on).

The pump and isolation valves are normally closed in a standby lineup (Green lights on).

Upon receipt of 901-4 H-15, the pump starts and the isolation valves open.

Distractor 1: Plausible if candidate does not recognize the Condensate pump isolation valves open and/or the Barometric pump running.

Distractor 2: Plausible if the candidate misinterprets Condensate Pump discharge lineup. When RCIC is running, this pump discharges to the RCIC Pump discharge header. If it discharged to the same place when shutdown, it would affect pump discharge and suction pressures.

Distractor 3: Plausible if the candidate misinterprets the inlet drain pot lineup.

Reference:

QCAN 901(2)-4 H-15 rev 6 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 217000 A3.06 Ability to monitor automatic operations of the Reactor Core Isolation Cooling System (RCIC) including:

Lights and alarms (RO=3.5 / SRO=3.4) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 80 of 154 29 February 2012 Associated objective(s):

217000.A3.06 Lights and alarms (RO=3.5 / SRO=3.4)

SR-1300-K06 (Freq: LIC=B)

Given a RCIC System 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.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 81 of 154 29 February 2012 40 ID: QDC.ILT.17126 Points: 1.00 An accident has occurred on Unit 1.

RPV water level is -70 inches and lowering slowly RPV pressure is 45 psig and steady The Unit Supervisor has given permission as stated in QGA 100: "OK to defeat the low RPV pressure and high area temperature isolations" (QCOP 1300-10, Bypassing RCIC Isolations:

LOW RPV Pressure or High Area Temperature).

The Field Supervisor has just reported that the necessary contacts have been blocked on relay 13A-K10 in the Aux Electric Room, bypassing the RCIC Low Pressure Isolation.

At Panel 901-4, which of the following operator action(s), if any, will be necessary to restart injection from RCIC?

A.

The RCIC Steam Supply Isolation Valves must be locally opened B.

The RCIC Isolation logic and Turbine must be reset C.

The 'RCIC MAN INITIATION' pushbutton must be pressed D.

No actions are necessary, RCIC will auto start and inject Answer:

B Answer Explanation:

Correct Answer: QCOP 1300-10 Step F.1.a blocks the necessary relay contacts and Step F.1.b directs operators to depress the STM LINE BRK TRIP RESET and the TURB RESET pushbuttons in order to reset the initiation logic. Since RPV water level is still less than -59 inches, RCIC will then auto start and inject.

Distractor 1: Plausible because the RCIC Steam Supply Isolation Valves must be opened manually, but can also be opened from the main control room.

Distractor 2: Plausible if the candidate believes RCIC must be re-initiated manually.

Distractor 3: Plausible if the candidate believes the turbine trip signal automatically resets as it does from a high reactor water level trip.

Reference:

QCOP 1300-10 rev 6 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 K/A:217000 Reactor Core Isolation Cooling System 2.4.20 Knowledge of the operational implications of EOP warnings, cautions, and notes.

(RO=3.8 / SRO=4.2) 10 CFR Part 55 Content: 41.7

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 82 of 154 29 February 2012 Question Source: New Question History: N/A Comments:

Associated objective(s):

217000.2.4.20 Knowledge of operational implications of EOP warnings, cautions, and notes. (RO=3.3 / SRO=4.0)

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

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 83 of 154 29 February 2012 41 ID: QDC.ILT.17128 Points: 1.00 Unit 1 was at full power when a complete loss of EHC Hydraulics occurred. Operators have successfully inserted a Manual Scram. The following conditions now exist:

All control rods are fully inserted RPV water level is 30 inches and steady RPV pressure is 885 psig and lowering Drywell pressure is 1.3 psig and steady Drywell temperature is 129ºF and rising slowly All ADS valves are open ADS Relief Valve Temperatures Panel 901-21 TR 1-0260-20 203-3A 323ºF 203-3B 322ºF 203-3C 321ºF 203-3D 323ºF 203-3E 321ºF 901-3 Panel Based on the above information:

(1) Are the listed Relief Valve temperatures as expected? Why or why not?

(2) Are the active alarms as expected?

A.

(1) NO; Relief Valve Tailpipe temperatures would show that of saturated steam at RPV pressure (~532ºF)

(2) NO; the Valve Leak Detection annunciator would NOT be active.

B.

(1) Yes; Relief Valve Tailpipe temperatures show that of super-heated steam at Torus pressure.

(2) NO; the Valve Leak Detection annunciator would NOT be active.

C.

(1) NO; Relief Valve Tailpipe temperatures would show that of saturated steam at RPV pressure (~532ºF)

(2) Yes; the active annunciators reflect the conditions of Relief Valves being open.

D.

(1) Yes; Relief Valve Tailpipe temperatures show that of super-heated steam at Torus pressure.

(2) Yes; the active annunciators reflect the conditions of Relief Valves being open.

Answer:

D

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 84 of 154 29 February 2012 Answer Explanation:

Tailpipe Temperatures rise to at least 323ºF in an isenthalpic process to a superheated state at Torus pressure, 1 psig/15 psia. Temps do not rise to saturation temp, the mistake made by operators during the accident at Three Mile Island.

The Valve Leak Detection annunciator (901-3 E-16) comes from TR 1-0260-20 on the 901-21 panel, which includes Drywell Cooler temps as well as Safety and Relief valve temps. (Note: This is also the reason that containment parameters are given as normal because elevated DW temps cause this alarm and mask the relief valve temp alarms otherwise.) Tailpipe temps above 300ºF cause this alarm.

h=1196.2 BTU/Lbm Saturated at 900 psia T-Sat =532ºF Superheated at 15 to 20 psia Tailpipe Temp= 323ºF Distractor 1: Combination of distractors 2 and 3.

Distractor 2: Plausible if the candidate assumes the high temperature alarm is based only on drywell temperature.

Distractor 3: Plausible if the candidate assumes the temperature of the steam exiting the relief valves is at normal reactor coolant temperature.

Reference:

QCAN 901-3 E-16, Steam Tables, INPO 88-008 Material For A Case Study On The Three Mile Island Unit 2 Accident Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 218000.A4.06 Ability to manually operate and/or monitor in the control room:

ADS valve tail pipe temperature (RO=3.5 / SRO=3.6) 10 CFR Part 55 Content: 41.3 Question Source: New Question History: N/A Associated objective(s):

218000.A4.06 ADS valve tail pipe temperature (RO=3.5 / SRO=3.6)

SR-0203-K20 (Freq: LIC=B)

Given various plant conditions, EVALUATE the following ADS valve indications /

responses and DETERMINE if the indication / response is expected and normal.

a. Electromatic relief valves/ PORVs/ Target Rock safety-relief indicating lights

b. Acoustic monitor indicating lights

c. ADS valve tail pipe temperature

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 85 of 154 29 February 2012 42 ID: QDC.ILT.17148 Points: 1.00 Unit 2 is recovering from an inadvertent Primary Containment Group (PCIS) Isolation with the following annunciators:

(NOT all alarms are listed) 902-5 A-8, GROUP 2 ISOL CH TRIP, is CLEAR (slow flash) 902-5 B-7, GROUP 1 ISOL CH TRIP, is CLEAR (slow flash) 902-5 B-4, GROUP 1 ISOL NOT RESET, is ALARMING (fast flash) 902-5 B-5, GROUP 2 ISOL NOT RESET, is ALARMING (fast flash)

If ONLY the 'ISOL VLV RESET' switch is placed in 'INBD' and then 'OUTBD', what will be the expected indications for the PCIS alarms on the 902-5 panel?

902-5 B-4, GROUP 1 ISOL NOT RESET will be (1).

902-5 B-5, GROUP 2 ISOL NOT RESET will be (2).

A.

(1) CLEAR (2) ALARMING B.

(1) ALARMING (2) CLEAR C.

(1) ALARMING (2) ALARMING D.

(1) CLEAR (2) CLEAR Answer:

B Answer Explanation:

The 'CH TRIP' alarms in slow flash indicate that the isolation signals for a Group 1 and 2 isolation are clear and the isolations can be reset.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 86 of 154 29 February 2012 The 'ISOL VLV RESET' switch will only reset a Group 2 and 3 isolation. A separate switch is required to reset the Group 1 isolation.

Distractor 1 is incorrect: Plausible because the Group 1 isolation switch is next to the ISOL VLV RESET switch on the 901-5 panel.

Distractor 2 is incorrect: Plausible if the candidate assumes that either both isolation switches are required to clear the isolations, or that conditions do not allow the isolations to be reset.

Distractor 3 is incorrect: 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-4 rev 7, QCAN 901(2)-5 B-5 rev 9 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 223002.A1.01 Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including:

System indicating lights and alarms (RO=3.5 / SRO=3.5) 10 CFR Part 55 Content: 41.7 Question Source: Modified from 2011 ILT NRC Exam Question History: N/A Comments:

Associated objective(s):

223002.A1.01 System indicating lights and alarms (RO=3.5 / SRO=3.5)

SR-1603-K10 (Freq: LIC=I) LIST the signals which cause the following Primary Containment Isolation (PCI) System isolations including purpose and setpoints.

DESCRIBE how they are bypassed AND how they are reset.

a. Group 1

b. Group 2

c. Group 3

d. Group 4

e. Group 5

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 87 of 154 29 February 2012 43 ID: QDC.ILT.17102 Points: 1.00 Following a scram on Unit 1, ERV 'B' repeatedly actuates to control reactor pressure.

What would the implications be if the vacuum breakers on the 'B' ERV tailpipe FAILED TO OPEN between actuations?

A.

ERV 'B' could fail to re-lift.

B.

ERV 'B' may NOT close after a subsequent lift.

C.

Damage to the ERV, piping and torus could occur.

D.

Higher cyclic containment stress due to higher pressure fluctuations.

Answer:

C Answer Explanation:

If both of the tailpipe vacuum breakers were stuck in the closed position and the corresponding relief valve was opened shortly after it had closed, overpressurization of relief valve piping and/or structural damage to the suppression pool could result due to water being drawn up into the tailpipe as the steam condenses.

Distractor 1 is incorrect: Plausible if the candidate assumed that the extra water in the tailpipe created too high a resistance for the relief valve to overcome on a subsequent relift.

Distractor 2 is incorrect: Plausible if the candidate assumed the vacuum created by the stuck closed vacuum breaker would be sufficient to prevent the relief valve from closing.

Distractor 3 is incorrect: The cyclic containment stress would actually be less with the vacuum breaker stuck closed because there would be less non-condensible gasses transferred to the torus airspace. Plausible if candidate assumes the pressure stresses introduced to the torus water would cause a significant rise in torus airspace pressure.

Reference:

LIC-0203 rev 17 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 239002.K5.06 Knowledge of the operational implications of the following concepts as they apply to RELIEF/SAFETY VALVES :

Vacuum breaker operation (RO=2.7 / SRO=3.0) 10 CFR Part 55 Content: 41.8 Question Source: Modified from 2008 Cooper ILT NRC Exam Question History: 2008 Cooper ILT NRC Exam Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 88 of 154 29 February 2012 Associated objective(s):

239002.K5.06 Vacuum breaker operation (RO=2.7 / SRO=3.0)

SR-0203-K22 (Freq: LIC=B)

Given various plant conditions, PREDICT how ADS valve and plant parameters will respond to the following ADS valve component or controller failures:

a. Target Rock bellows rupture

b. Tail pipe vacuum breaker stuck open/closed

c. ADS logic circuit power sensing relay failure

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 89 of 154 29 February 2012 44 ID: QDC.ILT.17139 Points: 1.00 Referring to the indications on the following page:

Predict the results of depressing the "CLOSE SLOW" pushbutton on the 'A' FRV M/A station for 2 seconds, and then releasing the pushbutton.

The 'A' FRV position will (1) and actual reactor water level will (2).

A.

(1) remain the same (2) remain the same B.

(1) close at 0.5% per second (2) remain the same C.

(1) close at 1% per second (2) lower and then settle at a new lower level D.

(1) close at 1% per second (2) lower and then return to the original level Answer:

A Answer Explanation:

The picture shows a screenshot of the "Overview" screen from the DFWLCS OWS, with the Unit at full power, all FRVs in Automatic, 3-Element, and in 'Group Control', the 'B' FRV selected as the primary, and the Second FRV Control in 'Sequence Manual.

The 'A' and 'B' FRV M/A station 'OPEN' and 'CLOSE' pushbuttons will only control FRV position if the FRV is in the 'Manual' mode of operation, and will have no effect in the

'Auto' mode.

This is the opposite of the very similar Reactor Recirc Pump individual control stations, that will change the speed of the pump if the operator depresses a manual speed change pushbutton with the master conroller in Master.

The candidate must recognize that each FRV control station must be individually selected to 'Manual' mode in order for the valve position control pushbuttons to have an effect.

Distractor 1 is incorrect: Plausible because 0.5% per second is the "Slow" rate of change for the Reactor Recirc pump manual control "Lower" pushbutton, and RPV water level will not change because the control station is in 'Auto'. This would be the correct answer if the question was asking the operation of the Recirc pump controller.

Distractor 2 is incorrect: Plausible because 1% per second is the correct rate of change for the 'CLOSE SLOW' pushbutton, however, RPV water level will not change because the control station is in 'Auto'. This would be the correct answer if the master FRV controller 'Lower' pushbutton were pressed.

Distractor 3 is incorrect: Plausible because 1% per second is the correct rate of change for the 'CLOSE SLOW' pushbutton, however, RPV water level will not change because the control station is in 'Auto'. This would be the correct answer if the 'A' FRV were in the

'Manual' position.

Reference:

LIC-0600 rev 8 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 90 of 154 29 February 2012 Tier: 2 Group: 1 K/A: 259002.A1.01 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR WATER LEVEL CONTROL SYSTEM controls including:

Reactor water level (RO=3.8 / SRO=3.8) 10 CFR Part 55 Content: 41.5 Question Source: New Question History: N/A Comments:

Associated objective(s):

SR-0600-K21 (Freq: LIC=B)

Given a Feedwater Level Control System operating mode and various plant conditions, PREDICT how feedwater level control/plant parameters will respond to manipulation of the following Feedwater Level Control System local/remote controls:

a. Low-Flow FWRV controller pushbuttons (1) Auto/Manual (2) Manual Adjust

b. A/B FWRV individual M/A transfer station pushbuttons (1) Auto/Manual (2) Manual Adjust

c. Feedwater Master Controller pushbuttons (1) 1 element/3 element selector (2) Automatic Setpoint Adjust

d. FWRV Lockup Reset pushbuttons

e. FWLC Operator Work Station (OWS)

(1) Process Overview Display controls (2) Measuring Points Display controls (3) Trend Display controls (4) Interlock Display controls (5) Alarm/Event Lists controls (6) OWS keyboard key (enable/disable) and keyboard

f. FWLC Hydraulic Skid local controls:

(1) Pump control switches (2) Nematron panel controls

g. Condensate recirc FCV 3401 controller (1) Auto/Manual (2) Manual Adjust

h. HS 1(2)-0640-362, U1(2) FWLC Power (ESS/INST) (2201(2)-190) 259002.A1.01 Reactor water level (RO=3.8 / SRO=3.8)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 91 of 154 29 February 2012 45 ID: QDC.ILT.17108 Points: 1.00 An unisolable steam leak to the Reactor Building (RB) has occurred on Unit 2, resulting in the RB being filled with steam.

RB Ventilation has isolated and Standby Gas Treatment (SBGTS) Train 'B' is running.

Steam is being drawn directly into the SBGTS suction.

Annunciator 912-7 A-5, STANDBY GAS B TRAIN TEMP HI, is in alarm The temperature downstream of the 'B' SBGTS Charcoal Adsorber is 300ºF and rising.

Based on the above conditions, this may result in (1), and requires the NSO to (2).

A.

(1) the charcoal catching fire (2) secure the 'B' SBGTS and start the 'A' SBGTS B.

(1) the charcoal catching fire (2) open the Fire spray isolation valve to the 'B' SBGTS C.

(1) the release of Iodine isotopes which had been adsorbed in the charcoal (2) secure the 'B' SBGTS and start the 'A' SBGTS D.

(1) the release of Iodine isotopes which had been adsorbed in the charcoal (2) open the Fire spray isolation valve to the 'B' SBGTS Answer:

C Answer Explanation:

The inclusion of steam into the suction of the SBGTS raises the moisture content of the charcoal. Charcoal has a higher affinity for water molecules than for Iodine. If water is present in the gas stream, the charcoal will release any Iodine molecules and bond with the water molecules. The adsorption of water by the charcoal also causes heat to be produced, and is the cause for the majority of the heat in the charcoal adsorber.

As temperature rises above 300ºF, this may also result in release of Iodine Isotopes.

The actions in QCAN 912-7 direct the 'B' SBGTS shut off and the 'A' SBGTS started.

Distractor 1 is incorrect: Plausible because the second part of the distractor is correct, and charcoal can catch fire if the temperature is high enough, but not until it reaches 650ºF Distractor 2 is incorrect: Combination of distractors 2 and 3.

Distractor 3 is incorrect: Plausible because the first part of the distractor is correct, and because a similar system does have a fire header valve for the purpose of putting out charcoal adsorber fires (CR HVAC).

Reference:

LF-7500 rev 17, QCAN 912-7 A-5 rev 4 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 92 of 154 29 February 2012 K/A: 261000.A2.04 Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

High train moisture content (RO=2.5 / SRO=2.7) 10 CFR Part 55 Content: 41.13 Question Source: New Question History: N/A Comments:

Associated objective(s):

SR-7500-K26 (Freq: LIC=B)

EVALUATE given key SBGTS 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 charcoal temperature

b. Failure to start automatically

c. Low or high system flow

d. Low or high system differential pressure 261000.A2.04 High train moisture content (RO=2.5 / SRO=2.7)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 93 of 154 29 February 2012 46 ID: QDC.ILT.17141 Points: 1.00 Which ONE of the following describes the effects (if any) on the 345 KV Switchyard GCBs on a complete loss of Relay House 125 VDC power?

The 345 KV Switchyard GCBs...

A.

CAN be operated from the Main Control Room AND the Relay House.

B.

CAN NOT be operated from the Main Control Room, but CAN be operated from the Relay House.

C.

CAN NOT be operated from the Relay House, but CAN be operated from the Main Control Room.

D.

CAN NOT be operated from the Main Control Room OR the Relay House.

Answer:

D Answer Explanation:

The Relay House 125 VDC system provides control power to both the trip and closing coils for each of the 345 KV Switchyard breakers. The ability to remotely operate the breakers from the Main Control Room and the Relay House will be lost. The breakers can only be operated locally in this condition.

Distractor 1 is incorrect: Plausible because the station 125 VDC system supplies control power to all other Main Control Room circuit breaker switches. The Lift Station 125 VDC system supplies control power for 13.8 KV breakers, in similar fashion to the Relay House 125 VDC to the 345 KV breakers.

Distractor 2 is incorrect: Plausible because the station 125 VDC system supplies control power to all other Main Control Room circuit breaker switches.

Distractor 3 is incorrect: Plausible because the Lift Station 125 VDC system supplies control power for 13.8 KV breakers, in similar fashion to the Relay House 125 VDC to the 345 KV breakers.

Reference:

LN-6100 rev 15 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 262001.K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the A.C. ELECTRICAL DISTRIBUTION:

D.C. power (RO=3.1 / SRO=3.4) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 94 of 154 29 February 2012 Associated objective(s):

SRN-6100-K23 (Freq: LIC=B NF=B)

Given a 345 KV Switchyard/ Main Transformer System operating mode and various plant conditions, PREDICT how the 345 KV Switchyard/ Main Transformer System will be impacted by the following support system failures:

a. Loss of MCC 18-2

b. Loss of 125 vdc

c. Loss of 480 vac

d. Loss of 120 vac

e. Fire protection deluge system actuation 262001.K6.01 D.C. power (RO=3.1 / SRO=3.4)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 95 of 154 29 February 2012 47 ID: QDC.ILT.17112 Points: 1.00 Unit 1 is at full power performing QCOS 6600-41, Unit 1 Emergency Diesel Generator Load Test.

DG Governor SPEED DROOP has been set to 50 The synchroscope is rotating at ONE (1) RPM in the 'FAST' (clockwise) direction The 'INCOMING VOLTS' meter is reading 121 AC Volts The 'RUNNING VOLTS' meter is reading 120 AC Volts The 'DIESEL GEN TO BUS 14-1 GCB' has just been closed at the 901-8 panel.

Which ONE of the following shows the correct indications based on the given conditions?

A.

B.

C.

D.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 96 of 154 29 February 2012 Answer:

A Answer Explanation:

The 'INCOMING VOLTS' meter reading higher than the 'RUNNING VOLTS' meter indicates that the EDG output voltage is higher than the voltage on Bus 14-1. When the EDG breaker is closed with a higher output voltage than Bus 13-1, KVARS will be positive.

The DG Speed Droop setting at 50 indicates that the DG is in Droop, vs Isochronous, and will therefore share load with Bus 13-1 based on the difference in frequency of the DG and the Main Generator. The synchroscope rotating at 1 revolution every 30 seconds in the fast direction indicates that the DG is operating at a slightly higher frequency than the Main Generator, and will therefore only pick up a small amount of load after the DG is paralleled.

Distractor 1 is incorrect: Plausible because this would be the correct answer if the Unit 1 EDG output voltage were lower than the Main Generator output voltage.

Distractor 2 is incorrect: Plausible because this would be the correct answer if the Unit 1 EDG were in Isochronous (Speed Droop at 0)

Distractor 3 is incorrect: Combination of distractors 1 and 2

Reference:

QCOS 6600-41 rev 40 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 262001.A1.03 Ability to predict and/or monitor changes in parameters associated with operating the A.C. ELECTRICAL DISTRIBUTION controls including:

Bus voltage (RO=2.9 / SRO=3.1) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 97 of 154 29 February 2012 Associated objective(s):

SRN-6500-K21 (Freq: LIC=B NF=B) Given a 4KV / 480 VAC Distribution Systems operating mode and various plant conditions, PREDICT how key system/plant parameters will respond to manipulation of the following 4KV / 480 VAC Distribution Systems local controls:

a. Transformer 11 control panels (1) CS-1 Fan Bank 1 control switch (2) CS-2 Second Stage Cooling control switch (3) Stage of Cooling control switch (43)

(4) Cooling fan breaker pushbuttons (5) Trouble alarm test switches (6) Sudden pressure relay cutout switch (on/off)

(7) T11 lockout relay reset switch

b. Transformer 21control panels (1) T21 Fan Bank 1 and 2 control switches and oil pumps control switch (3) Cooling fan and pump breakers (4) Trouble alarm test switches (5) Sudden pressure relay cutout switch (on/off)

(6) T21 lockout relay reset switch

c. Transformer control cabinets (Aux electric room)

(1) T12 / T22 lockout relay reset switch (2) T11/12/21/22 deluge reset pushbuttons

d. 4KV Buses (local)

(1) T12 (T22) UV relay reset pushbuttons on bus 12 and 14 (22 and 24)

(2) Bus 13-1,14-1 (23-1,24-1) second level undervoltage relay reset pushbuttons (3) Bus 13,14 (23,24) test switches (inhibit DG start)

e. 4KV breakers (local)

(1) Control switch (switch or pushbuttons)

(2) Test selector switch (3) 4KV breaker isolation/test switches (4) 4KV breaker 86 device reset switch (5) Charging motor cutout switch (6) Foot pedal (7) Local control box (pigtail)

(8) Bus 31 4KV grounding device knife switches (9) Vertical breaker racking direction selector switch and engage/start racking motor handle

f. 480 VAC breakers (local)

(1) Control switch (switch handle or pushbuttons)

(2) 480 VAC circuit breaker isolation/test switches (3) 480 VAC MCC breaker thermal reset (4) Local control box (pigtail)

g. Transformer 12/22 control panels (1) Lead cooler sequence switch (2) Auto mode enable test switch (3) Cooling fan breakers (4) LTC Raise/off/lower switch (43T-1)

(5) LTC Mode local/remote switch (43T-2)

(6) LTC Manual/off/auto switch (43T-3)

(7) Sudden pressure relay cutout switch (63 CO)

(8) Sudden pressure relay LTC and main tank off/enable switches (9) Lockout relay reset switch (cooling control)

(10) Trouble alarm test switches 262001.A1.03 Bus voltage (RO=2.9 / SRO=3.1)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 98 of 154 29 February 2012

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 99 of 154 29 February 2012 48 ID: QDC.ILT.17113 Points: 1.00 Unit 1 is in cold shutdown for a weekend maintenance outage. Electrical Maintenance has just finished performing an inspection of the ESS UPS Inverter. The current configuration is summarized below:

Annunciator 901-8 E-9, ESS SERV BUS ON EMERG SPLY, is in alarm and acknowledged (solid)

The ESS UPS is in a normal lineup and ready to accept load An EO has closed the ASCO Switch Normal Supply Breaker, supplying the output of the ESS UPS to the ESS ABT Which of the following describes the indications in the control room AFTER the EO places the ESS ABT 'TRANSFER CONTROL' switch in the 'RETRANSFER TO NORMAL' position?

A.

Annunciator 901-8 E-9, ESS SERV BUS ON EMERG SPLY, remains in alarm (solid)

B.

Annunciator 901-8 E-9, ESS SERV BUS ON EMERG SPLY, is NO LONGER in alarm (slow flash) ONLY C.

Annunciator 901-8 E-9, ESS SERV BUS ON EMERG SPLY, remains in alarm (solid)

AND 901-8 E-8, ESS SERV UPS ON DC OR ALT AC, is now in alarm (fast flash)

D.

Annunciator 901-8 E-9, ESS SERV BUS ON EMERG SPLY, is NO LONGER in alarm (slow flash)

AND 901-8 E-8, ESS SERV UPS ON DC OR ALT AC, is now in alarm (fast flash)

Answer:

B Answer Explanation:

The ESS ABT supplies power to the ESS Bus from either the Normal Supply (ESS UPS output), or the Emergency Supply (MCC 18-2).

'ESS SERV BUS ON EMERG SPLY' indicates that the ESS bus is currently being supplied by the Emergency supply (MCC 18-2) through the ESS ABT.

When the EO moves the the transfer switch on the ABT to 'RETRANSFER', the ESS Bus will be supplied power from the output of the ESS UPS, and the 901-8 E-9 alarm in the control room will clear.

Distractor 1 is incorrect: Plausible if the candidate assumed there are additional steps to complete the transfer.

Distractor 2 is incorrect: Plausible because the ESS Static Switch operation is very similar and the 901-8 E-8 alarm would come in if the Static Switch were selected to the alternate source.

Distractor 3 is incorrect: Plausible because this would be the correct answer if the ESS UPS output was aligned through the alternate AC supply.

Reference:

QCOP 6800-03 rev 32, QOA 900-8 E-9 rev 3 Reference provided during examination: None Cognitive level: High

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 100 of 154 29 February 2012 Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 262002.A4.01 Ability to manually operate and/or monitor in the control room:

Transfer from alternative source to preferred source (RO=2.8 / SRO=3.1) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A Comments:

Associated objective(s):

262002.A4.01 Transfer from alternative source to preferred source (RO=2.8 /

SRO=3.1)

SR-6800-K15 (Freq: LIC=I)

DESCRIBE the operation of the following principal Essential Service/Instrument Bus Systems components:

a. ESS Uninterruptable Power Supply (UPS)

b. ESS reserve AC ASCO ABT

c. Instrument bus ABT

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 101 of 154 29 February 2012 49 ID: QDC.ILT.17076 Points: 1.00 A plant transient on Unit 1 caused RCIC to receive an initiation signal.

RCIC is injecting at 400 GPM to the RPV.

The ANSO notes that all indications for RCIC are normal except for the MO 1-1301-17, STM SPLY ISOL VLV (OUTBD), which has NO lights lit on the 901-4 panel.

Which one of the following describes the cause of this indication?

(Assume NO operator actions have been taken)

A.

The valve has come off of its full open seat B.

A loss of AC power to the valve has occurred C.

A loss of DC power to the valve has occurred D.

A loss of 125 VDC control power to the valve has occurred Answer:

C Answer Explanation:

The power supply to the MO 1-1301-17 valve is 250 VDC RB MCC 1B.

Distractor 1 is incorrect: Plausible because other RCIC valves lose light indication when stroking.

Distractor 2 is incorrect: Plausible because the MO 1-1301-16 valve is AC powered.

Distractor 3 is incorrect: Plausible because the valve would lose light indication if it used 125 VDC control power. However, the valve is 250 VDC, and control power is provided directly from the 250 VDC Bus.

Reference:

QOM 1-6900-12 rev 5 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 263000.A3.01 Ability to monitor automatic operations of the D.C. ELECTRICAL DISTRIBUTION including:

Meters, dials, recorders, alarms, and indicating lights (RO=3.2 / SRO=3.3) 10 CFR Part 55 Content: 41.7 Question Source: Hope Creek ILT Exam Bank Question History: Hope Creek 2010 ILT NRC Exam Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 102 of 154 29 February 2012 Associated objective(s):

263000.A3.01 Meters, dials, recorders, alarms, and indicating lights (RO=3.2 /

SRO=3.3)

SRN-6900-K10 (Freq: LIC=I NF=I)

DESCRIBE how the Station DC Electrical Systems respond to a battery charger or DC bus/MCC breaker trip.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 103 of 154 29 February 2012 50 ID: QDC.ILT.15645 Points: 1.00 Unit 1 and Unit 2 are both at 100% power when an off-site grid disturbance initiates the following sequence of events:

0700: Unit 1 reactor scrams and T-12 trips 0721: Unit 2 reactor scrams and T-22 trips 0724: Unit 2 RPV water level is -65 inches 0730: Unit 1 RPV water level is -70 inches NO operator action has been taken. Complete the following statement:

At time 0731, Bus 13-1 is (1) and Bus 23-1 is (2).

A.

(1) de-energized (2) energized from the Unit 1/2 EDG B.

(1) de-energized (2) de-energized C.

(1) energized from the Unit 1/2 EDG (2) de-energized D.

(1) energized from the Unit 1/2 EDG (2) energized from the Unit 1/2 EDG Answer:

B Answer Explanation:

At 0700, the 1/2 EDG will auto-start and load to Unit 1 (Bus 13-1). At 0724, the Unit 1/2 EDG output breaker will open, Bus 13-1 will de-energize and the output breaker to Unit 2 (Bus 23-1) will close. The logic is designed so the 1/2 EDG will load onto the Unit that lost offsite power first. The LOCA signal on Unit 2 overrides this and trips the output breaker of the 1/2 EDG to Unit 1. At time 0730, both units now have a LOCA signal, and the 1/2 EDG output breakers will open. The 1/2 EDG will run unloaded until the operator places the desired unit's "1/2 DIESEL GEN OUTPUT GCB CONTROL" keylock switch to

'ON'.

Distractor 1 is incorrect: Plausible if candidate assumes that the EDG will load onto and stay on the unit that has the first LOCA signal.

Distractor 2 is incorrect: Plausible if the candidate does not realize that a LOCA signal will override a LOOP signal.

Distractor 3 is incorrect: Plausible because both Units require the 1/2 EDG and there is a procedure to supply Bus 13-1 and 23-1 from the 1/2 EDG. However, both 1/2 EDG output breakers will not auto-close on both ECCS buses.

Reference:

UFSAR Section 8.3.1.6.2. LN-6600 rev 18 page 51.

Reference provided during examination: N/A Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 1

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 104 of 154 29 February 2012 K/A: 264000.K3.02 Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following:

A.C. electrical distribution (RO=3.8 / SRO=4.2) 10 CFR Part 55 Content: 41.7 Question Source: Modified from Quad Cities 2009 ILT NRC Exam Question History: N/A Comments: None Associated objective(s):

264000.K3.02 A.C. electrical distribution (RO=3.8 / SRO=4.2)

SR-6600-K20 (Freq: LIC=B)

Given an Emergency Diesel Generators operating mode and various plant conditions, EVALUATE the following Emergency Diesel Generators indications/responses and DETERMINE if the indication/ response is expected and normal.

a. Diesel cooling water pump status

b. Generator volts, amps, frequency, kilowatts, and VARS

c. Output breaker position status

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 105 of 154 29 February 2012 51 ID: QDC.ILT.15684 Points: 1.00 Which one of the four Instrument Air Compressors (IAC) can ONLY be started from the Main Control Room and NOT from the local compressor control panel?

A.

1A IAC B.

1/2 IAC C.

1/2B IAC D.

U-2 IAC Answer:

B Answer Explanation:

The 1A, 1/2B, and U2 IAC control switches in the Main Control Room operate starter contactors on their respective compressors only. Their 480 VAC breakers must be closed locally during system startup. The operational effect of this is that three compressors can be started both locally and from the control room. The 1/2 IAC control switch in the Main Control Room operates both the starter contactor on the compressor and the closing coil for the compressor's power supply breaker at Bus 18. There is no option to start the 1/2 IAC from the local compressor control panel.

Distractors 1, 2, 3 are incorrect: Homogeneous distractors and the only other plausible answer options.

Reference:

QCOP 4700-10, Shutdown of an Instrument Air Compressor, Rev. 10. LN-4701 Instrument Air System, Rev. 10 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 KA: 300000.2.1.28 Knowledge of the purpose and function of major system components and controls. (RO=3.2 / SRO=3.3) 10 CFR Part 55 Content: 41.7 SRO Justification: N/A Question Source: New Question History: N/A Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 106 of 154 29 February 2012 Associated objective(s):

300000.2.1.28 Knowledge of the purpose and function of major system components and controls. (RO=3.2 / SRO=3.3)

SRN-4701-K16 (Freq: LIC=I NF=B) STATE the physical location and DESCRIBE the operation of the following Instrument Air System local controls:

a. 1A, 1/2, U2 Compressor control panel (1) Unload/Normal toggle (2) Reset/start (3) Stop

b. Dryer On/Off switch

c. Dryer Reset pushbutton

d. Prefilter manual blowdown toggle

e. Compressor control switches

f. 1/2B Electronikon controller

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 107 of 154 29 February 2012 52 ID: QDC.ILT.15689 Points: 1.00 Unit 1 was at rated power when a total loss of TBCCW occurs.

Which ONE of the following loads cooled by TBCCW has an automatic high temperature protection function?

A.

1A CRD Pump B.

1A Bus Duct Cooler C.

1A Service Air Compressor D.

1A Reactor Feed Pump Answer:

C Answer Explanation:

1A Service Air Compressor has a feature that trips the compressor on high outlet temperature of 450°F. All of the rest are major TBCCW loads but none of them have a high temperature trip feature.

Distractors 1, 2, & 3 are incorrect: Selected if any other load is identified as having a high temperature trip feature.

Reference:

QCOA 3800-03 rev 9, Total Loss of the TBCCW System Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 KA: 400000.K3.01 Knowledge of the effect that a loss or malfunction of the CCWS will have on the following:

Loads cooled by CCWS (RO=2.9 / SRO=3.3) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A Comments: None Associated objective(s):

400000.K3.01 Loads cooled by CCWS (RO=2.9 / SRO=3.3)

SRN-4600-K23 (Freq: LIC=B NF=B) Given a Service Air System operating mode and various plant conditions, PREDICT how the Service Air System will be impacted by the following support system failures: (Includes power supplies)

a. Loss of TBCCW

b. Loss of 480 vac

c. Loss of DC control power

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 108 of 154 29 February 2012

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 109 of 154 29 February 2012 53 ID: QDC.ILT.17053 Points: 1.00 Unit 1 has a LOCA in progress.

The Safe Shutdown Makeup Pump (SSMP) is the only source of injection.

The CCST levels are both 10,000 gallons and lowering slowly as SSMP injects.

What is the minimum action required to maintain the SSMP as an injection source for an indefinite period of time?

A.

Manually transfer suction to the fire header.

B.

Manually transfer suction to the service water header.

C.

Verify the auto transfer of the suction to the fire header.

D.

Verify the auto transfer of the suction to the service water header.

Answer:

A Answer Explanation:

SSMP draws a suction from either CCSTs or the fire header.

Distractor 1 is incorrect: Plausible because Service water provides cooling to the SSMP room coolers.

Distractor 2 is incorrect: Plausible because RCIC has an automatic supply transfer feature at 10,000 gallons CCST level.

Distractor 3 is incorrect: Combination of distractors 1 and 2.

Reference:

QCOP 2900-02 rev 23 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 1 K/A: 217000.K4.07 Knowledge of SAFE SHUTDOWN MAKEUP PUMP (SSMP) design feature(s) and/or interlocks which provide for the following:

Alternate supplies of water (RO=3.6 / SRO=3.6) 10 CFR Part 55 Content: 41.7 Question Source: Quad Cities ILT Exam Bank Question History: N/A Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 110 of 154 29 February 2012 Associated objective(s):

SR-2900-K26 (Freq: LIC=B)

EVALUATE given key SSMP 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. Low CCST level

b. Loss of service water

c. Loss of an AC power supply (4160/480/120)

d. Loss of 125 vdc 217000.K4.07 Alternate supplies of water (RO=3.6 / SRO=3.6)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 111 of 154 29 February 2012 54 ID: QDC.ILT.17054 Points: 1.00 A startup is in progress on Unit 1. Control rods are being withdrawn to raise the Flow Control Line to 50%.

The NSO reports annunciator 901-5 D-3, "TIMER MALFUNCT ROD SELECT BLOCK" alarms.

(1) How will this affect control rod movement?

(2) What action is required to restore the system and continue the startup?

A.

(1) ALL RMCS control rod motion is prevented.

(2) Place the standby Control Rod Sequence Timer in operation.

B.

(1) Control rod withdrawal ONLY is prevented.

(2) Bypass the RWM.

C.

(1) ALL RMCS control rod motion is prevented.

(2) Bypass the RWM.

D.

(1) Control rod withdrawal ONLY is prevented.

(2) Place the standby Control Rod Sequence Timer in operation.

Answer:

A Answer Explanation:

An RMCS Select Block prevents selection of any control rod by de-energizing the Rod Select relays; therefore, no rod movement can be accomplished.

Distractor 1 is incorrect: Incorrect because ALL rod motion via the RMCS system is prevented.

Distractors 2 and 3 are incorrect: Common Misconception due to similarity in the terms. RWM Select Blocks prevent selection of an out of sequence rod. Because of the annunciator given, it is clear that this is NOT an RWM block.

Reference:

QCAN 901(5) D-3, rev 5 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 201002.A3.01 Ability to monitor automatic operations of the REACTOR MANUAL CONTROL SYSTEM including:

Control rod block actuation (RO=3.2 / SRO=3.1) 10 CFR Part 55 Content: 41.6 Question Source: Quad Cities 2011 ILT Cert Exam Question History: Quad Cities 2011 ILT Cert Exam Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 112 of 154 29 February 2012 Associated objective(s):

SR-0280-K26 (Freq: LIC=B)

EVALUATE given key Reactor Manual Control System (RMCS)/ Rod Position Information System (RPIS) 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. Need to insert manual rod block

b. Rod withdrawal block

c. Rod select block

d. More than one rod selected

e. Control Rod Sequence Timer Failure 201002.A3.01 Control rod block actuation (RO=3.2 / SRO=3.1)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 113 of 154 29 February 2012 55 ID: QDC.ILT.17117 Points: 1.00 There is a latched Control Rod at a notch position where the reed switch has failed OPEN.

A substitute value has NOT been entered.

When looking at the Rod Worth Minimizer (RWM) display screen for that rod, how will its position be indicated?

A.

B.

++

C.

??

D.

Blank (nothing displayed)

Answer:

C Answer Explanation:

Rods with an unknown position from RPIS that do not have a substitute value will show

?? on the RWM display.

Distractor 1 is incorrect: Plausible because this would be the correct answer for rod at an intermediate (between notches) position.

Distractor 2 is incorrect: Plausible because this would be the correct answer for a fully withdrawn rod.

Distractor 3 is incorrect: Plausible because this is correct diplay description for the Full Core Display.

Reference:

LIC 0207 rev 13 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 201006.K6.03 Knowledge of the effect that a loss or malfunction of the following will have on the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) :

Rod position indication: P-Spec(Not-BWR6) (RO=2.9 / SRO=2.9) 10 CFR Part 55 Content: 41.6 Question Source: Hope Creek 2005 ILT NRC Exam Question History: Hope Creek 2005 ILT NRC Exam Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 114 of 154 29 February 2012 Associated objective(s):

SR-0207-K23 (Freq: LIC=B)

Given a Rod Worth Minimizer operating mode and various plant conditions, PREDICT how the Rod Worth Minimizer and supported systems will be impacted by the following failures:

a. Loss of single RPIS input signal

b. Loss of multiple RPIS input signals

c. Loss of FWLV inputs (steam/feed flow)

d. Loss of computer UPS

e. Rod drift

f. Mispositioned control rod 201006.K6.03 Rod position indication: P-Spec(Not-BWR6) (RO=2.9 / SRO=2.9)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 115 of 154 29 February 2012 56 ID: QDC.ILT.17056 Points: 1.00 Refer to the indications on the following page Unit 2 was operating at 100% power when the 2B Reactor Feed Pump tripped due to an electrical fault. One minute ago, the following alarm was received:

902-4 F-7, RECIRC LOOPS LIMITED BY FW FLOW/RX LVL Based on the above conditions, which of the following indications depicted will the operator see on the Individual Reactor Recirc Pump Speed Control Stations to ensure Recirc Flow is at the expected value?

A.

1 B.

2 C.

3 D.

4 Answer:

C Answer Explanation:

When the feed pump trips, a 45 second timer starts. If RWL reaches 26 inches and total steam flow is > 85% before the timer runs out, then the recirc pumps will run back to approximately 65% speed. 65% speed is about 70% core flow.

The presence of 902-4 F-7 indicates that all the requirements for the runback to occur have been met. One minute after the runback has occurred, the speed demand and actual speed of the recirc pumps will match within 0.1%.

Distractor 1 is incorrect: Plausible if candidate assumes that not enough time has passed for the recirc runback to occur.

Distractor 2 is incorrect: The runback is not to 70% speed. It is to a speed equivalent to 70% rated core flow.

Distractor 3 is incorrect: Plausible because there is a similar recirc runback to 32%

speed.

Reference:

QCAN 901(2)-4 F-7 rev 6 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 202002.A4.08 Ability to manually operate and/or monitor in the control room:

Recirculation system flow (RO=3.3 / SRO=3.3) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 116 of 154 29 February 2012 Comments: None Associated objective(s):

SR-0202-K06 (Freq: LIC=I)

Given a Reactor Recirculation System 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.

202002.A4.08 Recirculation system flow (RO=3.3 / SRO=3.3)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 117 of 154 29 February 2012 57 ID: QDC.ILT.17120 Points: 1.00 Unit 2 was at full power when a reactor scram occurred due to a spurious scram signal.

If QGA 100 were entered on RPV water level, what would be the expected condition of the RWCU system?

A.

The DEMIN BYP VLV MO 2-1201-133 is OPEN B.

The RWCU Recirc Pumps 2-1205A and 2-1205B are TRIPPED C.

The CU REJECT TO CONDENSER SV MO 2-1201-78 is OPEN D.

The PRECOAT PUMP 2-1279-7 has automatically started and is RUNNING Answer:

B Answer Explanation:

A scram from full power causes Reactor Water Level (RWL) to drop to approximately -10 inches. Zero inches is the QGA entry condition and also the Group 3 (RWCU) isolation signal. The Group 3 isolation causes the MO-1201-2, 5, and 80 valves to go shut, isolating the RWCU system. When the isolation valves go closed, the RWCU recirc pumps trip, the RWCU Filter Demins will go on hold, and the Hold pumps will start.

Distractor 1 is incorrect: Plausible if candidate assumes that the demin bypass valve opens to establish a flow path for the RWCU recirc pumps.

Distractor 2 is incorrect: Plausible if candidate assumes that the reject to condenser valve opens to establish a flow path for the RWCU recirc pumps.

Distractor 3 is incorrect: Plausible if candidate assumes the precoat pumps start to maintain precoat on the RWCU filter demins.

Reference:

QGA 100 rev 9, QCAN 901(2)-4 D-12 rev 10, QCAP 0200-10 rev 42 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 204000 Reactor Water Cleanup System 2.4.02 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (RO=4.5 / SRO=4.6) 10 CFR Part 55 Content: 41.7 Question Source: Modified from Quad Cities 2011 ILT NRC Exam Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 118 of 154 29 February 2012 Associated objective(s):

SR-1200-K11 (Freq: LIC=B)

LIST the signals which cause the following Reactor Water Cleanup System isolations including purpose and setpoints. DESCRIBE how they are bypassed AND how they are reset.

a. System isolation

b. Filter-demin isolation

c. Reject line isolation 204000.2.4.02 Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions. (RO=3.9 / SRO=4.1)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 119 of 154 29 February 2012 58 ID: QDC.ILT.17153 Points: 1.00 A fully withdrawn control rod that is subsequently inserted three-fourths of the way into the core will be displayed as position (1) and represents a rod travel of (2) feet into the core.

A.

(1) 12 (2) 3 B.

(1) 12 (2) 9 C.

(1) 36 (2) 3 D.

(1) 36 (2) 9 Answer:

B Answer Explanation:

There are 24 notch positions 6 inches apart between full out (48) and full in (00), equaling 12 feet of travel. A fully withdrawn rod inserted 3/4 of the way to full in would travel 9 ft.

and move from position 48 to position 12. There is no place to look up the answer to this postulated situation. A candidate must demonstrate understanding of CRDM operation.

Distractors 1, 2, & 3: Homogeneous distractors and possible misunderstandings of CRDM operation.

Reference:

LIC-0301 rev 10 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 214000.K1.03 Knowledge of the physical connections and/or cause-effect relationships between ROD POSITION INFORMATION SYSTEM and the following:

CRDM (RO=3.0 / SRO=3.1) 10 CFR Part 55 Content: 41.6 Question Source: New Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 120 of 154 29 February 2012 Associated objective(s):

214000.K1.03 CRDM (RO=3.0 / SRO=3.1)

SR-0301-K14 (Freq: LIC=I)

STATE the physical location and function of the following principal Control Rod Drive Blade / Mechanism components:

a. CRD Blade (1) Cruciform sheath (2) Absorber material (3) Velocity limiter (4) Alignment rollers (5) Female coupling socket (6) Coupling release handle (7) Male locking plug (8) Valve disc

b. CRD Mechanism (1) Insert and withdrawal ports (2) Ball-check valve (3) Outer tube (4) Inner tube (5) Index tube (6) Drive piston (a) Drive up/down seals (b) Inner/outer piston seals (c) Permanent ring magnet (7) Piston tube assembly (a) Indicator tube (b) Stop piston (c) Buffer holes (d) Bellville washers (8) Collet locking mechanism and Guide cap (a) Springs (b) Piston (c) Fingers (d) Guide cap (9) Drive screens (outer, inner, circumferential)

c. Assembly (1) Stub tube (2) Drive housing (3) Guide tube (4) Thermal sleeve

d. CRD housing support structure

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 121 of 154 29 February 2012 59 ID: QDC.ILT.17060 Points: 1.00 Unit 2 is at rated conditions performing QCOS 2300-05 Quarterly HPCI Pump Operability Test.

It is noted that on TR 2-1640-200B, TORUS H2O TEMP DIV 2, indicates the following:

'AVG TORUS H2O TEMP - POINT 9' reads 111.0ºF Based on the above information, which of the following is the HIGHEST importance action required?

NOTE: The responses below are listed in increasing order of importance A.

Enter QCAN 902-4 G-17, TORUS WTR HIGH TEMP B.

Enter QGA 200, PRIMARY CONTAINMENT CONTROL C.

Enter LCO 3.6.2.1 Suppression Pool Average Temperature, and Immediately suspend all testing that adds heat to the Torus D.

Enter LCO 3.6.2.1 Suppression Pool Average Temperature, and Immediately place the Mode Switch in SHUTDOWN Answer:

D Answer Explanation:

Torus water temperature is 111ºF. Therefore, the reactor must be immediately shutdown per TS 3.6.2.1 Condition D.

Distractor 1 is incorrect: The QCAN is entered at 90ºF torus temperature. However, TS 3.6.2.1 Condition D is more limiting and therefore takes precedence.

Distractor 2 is incorrect: QGA 200 is entered when Average Torus temperature is 95ºF.

However, TS 3.6.2.1 Condition D is more limiting and therefore takes precedence.

Distractor 3 is incorrect: Tech Specs require all testing that adds heat to the torus suspended if Torus temperature reaches 105ºF per TS 3.6.2.1 Condition C. However, Condition D is more limiting and therefore takes precedence.

Reference:

Tech Spec 3.6.2.1 Amendment no 199/195 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 223001.K5.11 Knowledge of the operational implications of the following concepts as they apply to PRIMARY CONTAINMENT SYSTEM AND AUXILIARIES :

Temperature measurement (RO=2.7 / SRO=2.7) 10 CFR Part 55 Content: 41.5 Question Source: Modified from Dresden 2006 ILT NRC Exam Question History: Dresden 2006 ILT NRC Exam Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 122 of 154 29 February 2012 Associated objective(s):

223001.K5.11 Temperature measurement (RO=2.7 / SRO=2.7)

SR-1601-K20 (Freq: LIC=B)

Given various plant conditions, EVALUATE the following Containment Systems indications/ responses and DETERMINE if the indication/ response is expected and normal.

a. Drywell pressure, air temperature, and water level

b. Drywell/torus differential pressure

c. Torus to Drywell vacuum breaker position

d. Torus pressure, air temperature, water level and temperature

e. Reactor building to torus vacuum breaker position

f. Reactor building temperatures (leak detection)

g. Reactor building/outside differential pressure

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 123 of 154 29 February 2012 60 ID: QDC.ILT.17135 Points: 1.00 Unit 1 was in Mode 2 with a startup in progress.

An EO in the Turbine Building reported a steam leak upstream of the Main Turbine Stop Valves and steam rapidly filling the Turbine Floor.

The following conditions are now present:

?

Reactor pressure is at 860 psig and lowering at 20 psig/min

?

RPV water level is 30 inches and steady

?

All Steam Line Flow indicators are downscale

?

MSIV Room temperatures are stable Given the listed conditions, over the next 10 minutes:

The MSIVs (1) close automatically due to a Group 1 Isolation.

The Reactor (2) procedurally required to be scrammed prior to the MSIVs being closed.

A.

(1) WILL (2) IS B.

(1) WILL (2) IS NOT C.

(1) WILL NOT (2) IS D.

(1) WILL NOT (2) IS NOT Answer:

C Answer Explanation:

Based on the given trend, Reactor pressure will lower to the Group 1 initiation setpoint (785 psig) in the next 10 minutes. However the Group 1 Isolation is not armed with the Mode Switch not in Run, as given by the stem with the Reactor in Mode 2. The other Group 1 Isolation signals: High Steam tunnel temp, High Steam line flow, and Low RPV level, are all given as stable. Therefore, an automatic closure of the MSIVs will not occur.

For a steam leak outside the containment at an unknown location, QCOA 0201-05, Primary System Leaks Outside Primary Containment, directs operators to manually scram and initiate a Group 1 Isolation.

Distractor 1 is incorrect: Plausible if the candidate does not recognize the Group 1 is not armed with the given conditions.

Distractor 2 is incorrect: Plausible if the candidate does not recall the direction to manually insert a reactor scram and then isolate the leak by closing the MSIVs.

Distractor 3 is incorrect: Plausible if the candidate recognizes the Group 1 is not armed, but does not recall the direction to scram the reactor prior to closing the MSIVs.

Reference:

QCOA 0201-05 rev 10, LN-1603 rev 11 Reference provided during examination: None Cognitive level: High

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 124 of 154 29 February 2012 Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 239001.A2.11 Ability to (a) predict the impacts of the following on the MAIN AND REHEAT STEAM SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Steam line break (RO=4.1 / SRO=4.3) 10 CFR Part 55 Content: 41.10 Question Source: New Question History: N/A Comments:

Associated objective(s):

239001.A2.11 Steam line break (RO=4.1 / SRO=4.3)

SR-1603-K10 (Freq: LIC=I) LIST the signals which cause the following Primary Containment Isolation (PCI) System isolations including purpose and setpoints.

DESCRIBE how they are bypassed AND how they are reset.

a. Group 1

b. Group 2

c. Group 3

d. Group 4

e. Group 5

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 125 of 154 29 February 2012 61 ID: QDC.ILT.17062 Points: 1.00 Which of the following describes the response of the Main Turbine Control Valves (CVs) AND Intercept Valves (IVs) as Main Turbine speed rises from 100% to 107% rated speed?

A.

The CVs AND IVs fast close together B.

CVs throttle closed first followed by the IVs C.

IVs throttle closed first followed by the CVs D.

CVs throttle closed AND the IVs fast close Answer:

B Answer Explanation:

Control and Intercept valves control turbine speed during overspeed conditions. The CVs close first to regulate speed followed by the IVs.

Distractor 1 is incorrect: The CVs and IVs will not fast close together on an overspeed condition. This would be the correct answer if a Power Load Imbalance condition were to occur.

Distractor 2 is incorrect: The CVs close first, not the IVs.

Distractor 3 is incorrect: The CVs will throttle closed, however the IVs do not fast close.

Reference:

LIC-5652a rev 5 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 241000.A1.13 Ability to predict and/or monitor changes in parameters associated with operating the REACTOR/TURBINE PRESSURE REGULATING SYSTEM controls including:

Main turbine speed (RO=2.7 / SRO=2.7) 10 CFR Part 55 Content: 41.5 Question Source: Nine Mile Point 2002 ILT NRC Exam Question History: Nine Mile Point 2002 ILT NRC Exam Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 126 of 154 29 February 2012 Associated objective(s):

SR-5652a-K22 (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 the following failures:

a. Pressure regulator fails high or low

b. Load reject

c. Turbine trip

d. Overspeed

e. Reactor scram

f. Loss of stator cooling 241000.A1.13 Main turbine speed (RO=2.7 / SRO=2.7)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 127 of 154 29 February 2012 62 ID: QDC.ILT.17064 Points: 1.00 Unit 2 is operating at 50% Reactor power.

A disturbance on the grid results in the following indications:

"ALARM S1_P2 POWER LOAD UNBALANCE OCCURRED" is in ALARM on the DEHC HMI 902-7 D-15, POWER LOAD UNBALANCE TRIP, is in ALARM Which one of the following describes how the DEHC system INITIALLY responds to these conditions?

A.

The Turbine Control Valves AND Intercept Valves FAST CLOSE, and the Bypass Valves OPEN B.

The Bypass Valves OPEN ONLY C.

Turbine Control Valves CLOSE slowly to reduce stator amps D.

The Turbine Stop Valves CLOSE and the Bypass Valves OPEN Answer:

A Answer Explanation:

The Power Load Unbalance (PLU) function of DEHC is activated by the electrical load reject condition and will cause the Turbine Control Valves (TCVs) to Fast Close as well as the Intercept valves. The resultant rise in pressure will cause the Bypass Valves to open to control reactor pressure.

Distractor 1 is incorrect: Plausible if the candidate assumes the BPVs will open only to relieve load from the turbine.

Distractor 2 is incorrect: Plausible because the Turbine will runback (reduce load) until stator amps are less than 9121 amps on a loss of stator cooling.

Distractor 3 is incorrect: Plausible if the candidate assumes the Turbine will trip directly from a Power Load Unbalance condition. The reactor will scram directly from the PLU signal, which will eventually cause the Main Generator to trip on reverse power, tripping the turbine.

Reference:

QOA 900-7 D-15 rev 2, QOA 900-5 H-4 rev 9, QCOA 5650-04 rev 9 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 245000.K4.09 Knowledge of MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS design feature(s) and/or interlocks which provide for the following: Turbine control (RO=3.1 / SRO=3.2) 10 CFR Part 55 Content: 41.5 Question Source: New Question History: N/A

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 128 of 154 29 February 2012 Comments:

Associated objective(s):

SR-5652a-K22 (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 the following failures:

a. Pressure regulator fails high or low

b. Load reject

c. Turbine trip

d. Overspeed

e. Reactor scram

f. Loss of stator cooling 245000.K4.09 Turbine control (RO=3.1 / SRO=3.2)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 129 of 154 29 February 2012 63 ID: QDC.ILT.17067 Points: 1.00 Which of the following Fire System components will be affected by a loss of 125 VDC Reactor Building Bus 1?

A.

The FAS (Fire Alarm System) Control Panel B.

Unit 1 HPCI Protectowire system C.

Service Water Supply Valve, MO-1/2-3906 D.

The Fire Diesel Battery Charger Answer:

B Answer Explanation:

Power supply to the Unit 1 HPCI Protectowire system is from 125 VDC Reactor Building bus 1.

Distractor 1 is incorrect: Power supply to the FAS terminal is from the Computer UPS with a backup from an internal battery.

Distractor 2 is incorrect: Power supply to the 1/2 3906 valve is from 250 VDC MCC 1.

Distractor 3 is incorrect: Power supply to Fire Diesel Battery Charger is from 120 VAC Distribution panel 16-2-1.

Reference:

LN-4100 rev 21, QOA 6900-12 rev 16 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 286000.K2.03 Knowledge of electrical power supplies to the following:

Fire detection system: Plant-Specific (RO=2.5 / SRO=2.7) 10 CFR Part 55 Content: 41.7 Question Source: New Question History: N/A Comments: None Associated objective(s):

SRN-4100-K19 (Freq: LIC=I NF=I)

LIST the plant systems which support Fire Protection Systems and DESCRIBE the nature of support. (Includes power supplies) 286000.K2 03 Fire detection system: Plant-Specific (RO=2.5 / SRO=2.7)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 130 of 154 29 February 2012

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 131 of 154 29 February 2012 64 ID: QDC.ILT.17122 Points: 1.00 Unit 1 is operating at 100% power. The following conditions exist:

Annunciator 901-3 E-3, RX BLDG VENT CHANNEL A DOWNSCALE, is LIT Annunciator 901-3 F-3, RX BLDG VENT CHANNEL B DOWNSCALE, is LIT The ANSO checks the status of the Reactor Building (RB) vents and finds the vent fans running and the isolation dampers open with 1/2 B Standby Gas Treatment System (SBGTS) train running.

Based ONLY on the above conditions, what actions, if any, are required and why?

A.

STOP the 1/2 B SBGTS train because running it in conjunction with the RB vents may cause damage to the train.

B.

TRIP the RB supply and exhaust fans and CLOSE the isolation dampers to prevent an unfiltered release of radioactive particles.

C.

Leave both systems RUNNING to ensure the Secondary Containment pressure remains negative.

D.

TRIP the RB supply and exhaust fans and leave the isolation dampers OPEN to ensure all effluent is processed through SBGTS.

Answer:

B Answer Explanation:

Both Reactor Building Vent Channels downscale is an initiation signal for Reactor Building ventilation to isolate and Standby Gas Treatment System to start. The stem indicates that while SBGTS started, the isolation of Reactor Building ventilation failed to occur.

Distractor 1 is incorrect: The SBGTS should be left running because of the initiation signal present.

Distractor 2 is incorrect: The Reactor Building ventilation system should be isolated because it exhausts unfiltered, non-elevated air. SBGTS is enough to maintain the Reactor Building differential pressure negative.

Distractor 3 is incorrect: Leaving the Reactor Building ventilation system dampers open creates an additional inflow of air which may not allow the SBGTS to maintain a negative Reactor Building different pressure.

Reference:

QCAN 901(2)-3 E-3 rev 5, QCAN 901(2)-3 F-3 rev 5 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 290001.K3.01 Knowledge of the effect that a loss or malfunction of the SECONDARY CONTAINMENT will have on following:

Off-site radioactive release rates (RO=4.0 / SRO=4.4)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 132 of 154 29 February 2012 10 CFR Part 55 Content: 41.9 Question Source: Bank Question History: Quad Cities 2003 ILT NRC Exam Comments:

Associated objective(s):

290001.K3.01 Off-site radioactive release rates (RO=4.0 / SRO=4.4)

SR-5750-K20 (Freq: LIC=B) Given a Plant Ventilation Systems operating mode and various plant conditions, EVALUATE the following Plant Ventilation Sstems indications/responses and DETERMINE if the indication/ response is expected and normal.

a. Reactor building ventilation (1) Differential pressures (2) Damper positions (3) Building supply/exhaust fan status and amperage (4) Supply, exhaust and outside air temperatures

b. Turbine building ventilation (1) Differential pressures (2) Damper positions (3) Building supply/exhaust fan status and amperage (4) Main chimney flow rate (5) East/west supply, and exhaust air temperatures

c. Radwaste building ventilation:

(1) Building supply/exhaust fan and damper status (2) Differential pressures, temperatures

d. Off-gas filter building ventilation (1) Fan status and damper positions (2) Differential pressures, temperatures and flow (3) Digital Mimic Display on Panel 2212-37B(A)

(a) Flashing Red Light (computer hardware failure)

e. Off-gas filter building freon leak detector indication

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 133 of 154 29 February 2012 65 ID: QDC.ILT.17097 Points: 1.00 Unit 1 was operating at full power when the Control Room Ventilation System isolated due to High Main Steam Line flow.

After the alarm condition clears, the isolation must be reset at the...

A.

901-5 panel OR 912-5 panel ONLY.

B.

901-5 panel AND 1/2-9400-105 panel ONLY.

C.

912-5 panel AND 1/2-9400-105 panel ONLY.

D.

901-5, 912-5 AND 1/2-9400-105 panels.

Answer:

B Answer Explanation:

An isolation of the CRHVAC system due to High Main Steam Line flow will need to be reset at both the local control panel (1/2-9400-105 panel) and in the control room to reset the locked in CRHVAC isolation signal (901-5 panel).

Distractor 1 is incorrect: Plausible because the signal must be reset at the 901-5 panel and because an isolation signal can be initiated from the 912-5 panel, but not cleared.

Distractor 2 is incorrect: Plausible because the signal must be reset at the 1/2-9400-105 panel and because an isolation signal can be initiated from the 912-5 panel, but not cleared.

Distractor 3 is incorrect: Combination of distractors 2 and 3.

Reference:

QOA 900-5 D-8 rev 2, LN-5752 rev 14 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 2 Group: 2 K/A: 290003.A4.01 Ability to manually operate and/or monitor in the control room:

Initiate/reset system (RO=3.2 / SRO=3.2) 10 CFR Part 55 Content: 41.9 Question Source: New Question History: N/A Comments:

Associated objective(s):

290003.A4.01 Initiate/reset system (RO=3.2 / SRO=3.2)

SR-5752-K07 (Freq: LIC=I)

LIST the signals which cause a Control Room Ventilation System auto initiation including setpoints. DESCRIBE how they are bypassed AND how they are reset.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 134 of 154 29 February 2012

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 135 of 154 29 February 2012 66 ID: QDC.ILT.17131 Points: 1.00 Given:

The plant is 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.

Distractors 1, 2 and 3 are incorrect: 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 K/A: 2.1.01 Knowledge of conduct of operations requirements. (RO=3.8 / SRO=4.2) 10 CFR Part 55 Content: 41.10 Question Source: Quad Cities ILT Exam Bank Question History: Quad Cities 2011 ILT NRC Exam Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 136 of 154 29 February 2012 Associated objective(s):

2.1.01 Knowledge of conduct of operations requirements. (RO=3.8 / SRO=4.2)

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).

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 137 of 154 29 February 2012 67 ID: QDC.ILT.17132 Points: 1.00 Unit 2 was at 10% reactor power, performing a normal reactor startup, when the following sequence of events occurred:

A control rod was selected to be withdrawn from position '00' to position '12' Because the control rod was difficult to move from position '00', the Unit Supervisor authorized the use of "Double Clutching" IAW QCOA 0300-02, Inability To Drive A Control Rod Stuck The NSO performed the double clutching procedure and immediately released the 'ROD MOVEMENT CONTROL switch' and 'ROD OUT NOTCH OVERRIDE switch' when the rod started to move The rod settled at position 06 In accordance with QCGP 4-1, Control Rod Movements and Control Rod Sequence:

(1) Did the actions to move the control rod from position '00' comply with the requirements for the use of the continuous control rod withdrawal feature?

(2) Which method(s) may be used to withdraw the control rod to position '12'?

A.

(1) NO, the requirements were NOT met.

(2) Single notch to position '12' ONLY B.

(1) NO, the requirements were NOT met.

(2) Continuous rod withdrawal to position '10', then single notch to position '12' C.

(1) Yes, the requirements were met.

(2) Single notch to position '12' ONLY D.

(1) Yes, the requirements were met.

(2) Continuous rod withdrawal to position '10', then single notch to position '12' Answer:

C Answer Explanation:

The Operations Philosophy Handbook section on Conservative Decision Making requires that plant procedures incorporate steps for reactivity changes only in a deliberate, carefully controlled manner. That requirement is reflected in QCGP 4-1 Step E.18. The note before step E18.b allows use of the continuous control rod withdrawal feature for rods that are difficult to move from Full-In. Section E.18.b dis-allows continuous rod movement, either in or out, for rod movement of 3 notches or less.

Distractor 1: Plausible because the rod moved 3 notches in continuous override, which is not normally allowed. However, the limitiation for using continuous rod withdrawal does not apply when a rod is starting from the full in position.

Distractor 2: Combination of distractors 1 and 3 explanations.

Distractor 3: Plausible if the candidate assumes the limitation for continuous use applies to control rod positions, instead of notches, in which the control rod would move more than 3 positions (6 to 10).

Reference:

QCGP 4-1 Control Rod Movements and Control Rod Sequence, Rev. 39, Step E.18 OP-AA-101-111-1001, Operations Philosophy Handbook, Rev. 9, Step 4.11.6

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 138 of 154 29 February 2012 QCOA 0300-02, Inability to Drive a Control Rod: Control Rod Stuck, Rev. 17 Reference provided during examination: None Cognitive level: High Level (RO/SRO): RO Tier: 3 KA Number and Statement: 2.1.39 (RO=3.6 / SRO=4.3)

Knowledge of conservative decision making practices 10 CFR Part 55 Content: 41.10 Question Source: New Question History: N/A Comments:

Associated objective(s):

SR-0002-P01 (Freq: LIC=B)

Given a reactor plant during a startup, perform a reactor startup consisting of the following tasks in accordance with QCGP 1-1:

a. Criticality and establish a heatup

b. Transfer mode switch to run

c. Turbine roll and synchronization 2.1.39 Knowledge of conservative decision making practices. (RO=3.6 / SRO=4.3)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 139 of 154 29 February 2012 68 ID: QDC.ILT.15656 Points: 1.00 Given:

Unit 2 is at rated power.

Drywell temperature sensor 2-5741-43A has failed.

All other temperature sensors in the same area are reading correctly.

The failed sensor feeds Point 8 on recorder 2-2340-9, DRYWELL TEMPERATURE RECORDER at the 902-3 panel and also inputs to Annunciator 902-3 H-4, DRYWELL HIGH AIR TEMPERATURE.

Which of the following methods is used to track and identify the inoperable alarm point?

A.

Affix a TCC Tag to the defective indicator per CC-AA-112, Temporary Configuration Changes.

B.

Install a temporary label on the defective indicator per OP-AA-116-101, Equipment Labeling.

C.

Post an Information Tag adjacent to the defective indicator per OP-AA-109-101, Clearance and Tagging.

D.

Place an Equipment Deficiency Tag adjacent to the defective indicator per OP-AA-108-105, Equipment Identification and Documentation.

Answer:

D Answer Explanation:

Answer: The failed temperature indicator is an example of a Main Control Room Deficiency from attachment 5 of OP-AA-108-105. An Equipment Deficiency Tag is generated and placed adjacent to the defective indicator per OP-AA-108-105-1001.

Distractor 1: CC-AA-112 would be used if the alarm were disabled by the operator or if a temporary modification were to be put in place to monitor the indication.

Distractor 2: A temporary label would only be used if the equipment was new and didn't yet have a label or the label was changed or damaged.

Distractor 3: OP-AA-109-101, Clearance and Tagging, is the process used to protect personnel while performing work on systems. It is designed for protection, prevention of inadvertent operation, and administrative control when necessary.

Reference:

OP-AA-108-105 rev 8, OP-AA-108-105-1001 rev 5, QCAN 902-3 H-4 rev 4.

Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: 1 K/A: 2.2.43 Knowledge of the process used to track inoperable alarms. (RO=3.0 /

SRO=3.3)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 140 of 154 29 February 2012 10 CFR Part 55 Content: 41.10 Question Source: New Question History: N/A Comments: None Associated objective(s):

2.2.43 Knowledge of the process used to track inoperable alarms. (RO=3.0 / SRO=3.3)

SRNLF-CO-K3 (Freq: LIC=B NF=B)

DESCRIBE and STATE the purpose of each type of card used in OP-MW-109-101, Clearance and Tagging.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 141 of 154 29 February 2012 69 ID: QDC.ILT.15652 Points: 1.00 From the following, select which statement describes "Operable-Operability", in accordance with Quad Cities Technical Specifications.

The condition of a system, subsystem, division, component, or device...

A.

necessary to protect the integrity of certain physical barriers to guard against the uncontrolled release of radioactivity.

B.

capable of performing its specified safety function(s) independent of its support systems.

C.

that will allow testing, calibration or inspection to assure operation is within Safety Limits and LCOs.

D.

capable of performing its specified safety function(s) with its support systems capable of performing their required support function(s).

Answer:

D Answer Explanation:

Per T.S. section 1.1 Definitions: A system, subsystem, division, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified safety function(s) and when all necessary attendant instrumentation, controls, normal or emergency electrical power, cooling and seal water, lubrication, and other auxiliary equipment that are required for the system, subsystem, division, component, or device to perform its specified safety function(s) are also capable of performing their related support function(s).

Distractor 1: This is the definition of a Safety Limit Distractor 2: It IS required to have all its support systems capable of performing their required support functions.

Distractor 3: This is the definition of a Surveillance.

Reference:

Tech Specs Section 1.1 Definitions Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 K/A: 2.2.38 Knowledge of conditions and limitations in the facility license. (RO=3.6 /

SRO=4.5) 10 CFR Part 55 Content: 41.7 Question Source: LaSalle ILT Exam Bank Question History: LaSalle 2008 ILT NRC Exam Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 142 of 154 29 February 2012 Associated objective(s):

2.2.38 Knowledge of conditions and limitations in the facility license. (RO=3.6 /

SRO=4.5)

SR-OPDT-K02 (Freq: LIC=A) DEFINE the following words or phrases associated with Operability Determination.

a. Compensatory Measure

b. Consequential Failure

c. Current Licensing Basis (CLB)

d. Degraded Condition

e. Design Basis

f. Functional

g. Nonconforming Condition

h. Operable

i. Operability Determination

j. Operability Evaluation

k. Reasonable Expectation

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 143 of 154 29 February 2012 70 ID: QDC.ILT.15660 Points: 1.00 Given the Technical Specification Surveillance requirement shown below:

The channel check was last performed on 09/01 at 0700.

What is the LATEST time the next channel check must be completed?

A.

09/01 at 1900 B.

09/01 at 2200 C.

09/02 at 0700 D.

09/02 at 1900 Answer:

B Answer Explanation:

SR 3.0.2 states " The specified frequency for each SR is met if the Surveillance is performed within 1.25 times the interval specified in the Frequency, as measured from the previous performance or as measured from the time a specified condition of the Frequency is met".

Since the Frequency is NOT specified as "once", it qualifies for the 1.25 extension.

The correct answer is 1.25 X 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> = 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> from the initial performance. The surveillance is due on 09/01 at 2200.

Distractor 1: Adds the surveillance frequency of 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

Distractor 2: Adds 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> from original performance at 09/01 at 0700.

Distractor 3: Adds 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the initial 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> frequency.

Reference:

Tech Specs section 1.4 Frequency, Amendment No. 223/218 Reference provided during examination: N/A Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 144 of 154 29 February 2012 K/A: 2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (RO=3.9 / SRO=4.6) 10 CFR Part 55 Content: 41.7 Question Source: Quad Cities ILT Exam Bank Question History: N/A Comments:

Associated objective(s):

2.2.42 Ability to recognize system parameters that are entry-level conditions for Technical Specifications. (RO=3.9 / SRO=4.6)

SRNLF-00-K08 (Freq: LIC=I NF=I)

Given symptoms and indications depicting a generic abnormal condition and the administrative procedures, DESCRIBE the operator actions of the applicable administrative procedure.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 145 of 154 29 February 2012 71 ID: QDC.ILT.15664 Points: 1.00 Which of the following identifies the lowest off-site release rate that, if exceeded, requires entry into QGA-400, Radioactivity Release Control?

Emergency Action Level...

A.

'Unusual Event' (4.74 E+04 µCi/sec)

B.

'Alert' (8.34 E+05 µCi/sec)

C.

'Site Area Emergency' (1.62 E+06 µCi/sec)

D.

'General Emergency' (1.62 E+07 µCi/sec)

Answer:

B Answer Explanation:

>8.34 E+05 µCi/sec is the ALERT EAL classification threshlold value and is the entry condition into QGA 400.

Distractor 1: 4.74 E+04 µCi/sec is the Unusual Event EAL classification threshold value and is below the value for entry into QGA 400.

Distractor 2: 1.62 E+06 µCi/sec is the Site Area Emergency EAL classification threshold value and is above the value for entry into QGA 400.

Distractor 3: 1.62 E+07 µCi/sec is the General Emergency EAL classification threshold value and is above the value for entry into QGA 400.

Reference:

QGA 400 Rev 7, EP-AA-1006 Rev 31 Reference provided during examination: N/A Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A K/A: 2.3.11 Ability to control radiation releases. (RO=3.8 / SRO=4.3) 10 CFR Part 55 Content: 41.12 Question Source: Modified from LaSalle 07-01 NRC ILT Exam Question History: N/A Comments: None Associated objective(s):

SR-0001-K33 (Freq: LIC=B)

STATE the entry conditions to QGA 400, 'Radioactivity Release Control'.

2.3.11 Ability to control radiation releases. (RO=3.8 / SRO=4.3)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 146 of 154 29 February 2012

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 147 of 154 29 February 2012 72 ID: QDC.ILT.15650 Points: 1.00 Complete the following statement that correctly identifies an activity that would allow Independent Verification to be WAIVED?

Removing a Danger Tag from...

A.

1-1101-4 (SBLC TK OUTLET VLV) while in Hot Shutdown.

B.

1-3599-50 (1B3 HTR NORM LCV INLET VLV) at rated plant conditions.

C.

1-1002A RHR PUMP 1A BKR (BUS 13-1 CUBICLE 9) while in Hot Shutdown.

D.

MO 1-3703 BKR (U1 RBCCW OUTBOARD RETURN VLV) at rated plant conditions.

Answer:

B Answer Explanation:

The Shift Manager may WAIVE verification requirements for ALARA concerns. This valve is not safety related and is located in the Low Pressure Heater Bay. This room is a locked high radiation area and radiation levels would be high at rated conditions.

Distractor 1: Independent Verification (IV) shall be performed when removing danger tags from equipment governed by tech specs. SBLC is required to be operable per tech specs in Mode 3.

Distractor 2: IV shall be performed when removing danger tags from equipment governed by tech specs. RHR is required to be operable per tech specs in Mode 3.

Distractor 3: IV shall be performed when removing danger tags from safety related equipment. The MO 1-3703 valve is a primary containment isolation valve and is safety related equipment.

Reference:

HU-AA-101 rev 5 Reference provided during examination: None Cognitive level: Memory Level (RO/SRO): RO Tier: 3 Group: N/A K/A: 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) 10 CFR Part 55 Content: 41.12 Question Source: Modified from Monticello 2009 ILT NRC Exam Question History: N/A Comments: None

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 148 of 154 29 February 2012 Associated objective(s):

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)

SRNLF-PGH-K3 (Freq: LIC=B NF=B)

From memory, DESCRIBE the following Human Performance Tools and Verification Practices in accordance with HU-AA-101 and OP-AA-104-101:

a. Self Check (STAR)

b. Outside Procedures, Parameters, or Processes (OOPS)

c. Peer Check

d. Independent Verification

e. Concurrent Verification

f. Alternate Verification Techniques

f. 3 Way Communication

g. First Check

h. Flagging/Robust Operational Barriers

i. Two Minute Drill

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 149 of 154 29 February 2012 73 ID: QDC.ILT.15651 Points: 1.00 Which of the following correctly prioritizes the following procedures and conditions?

Consider each procedure/condition separately.

1. QGA 200 for High Torus Water Level

2. QCOA 3800-01 for total loss of TBCCW

3. SAMGs with TSC ready to assume command and control and RPV level less than -191 inches and can NOT be restored following Emergency Depressurization A.

1, 2, 3 B.

1, 3, 2 C.

2, 3, 1 D.

3, 1, 2 Answer:

D Answer Explanation:

The hierarchy of procedure execution is SAMGs, QGAs, QCOA/QOAs, and QCOP/QOPs.

Distractors 1,2 and 3: Homogenous distractors.

Reference:

L-QGAINT rev 7 Reference provided during examination: N/A Cognitive level: High Level (RO/SRO): RO Tier: 3 Group: N/A K/A: 2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. (RO=3.5 / SRO=4.4) 10 CFR Part 55 Content: 41.10 Question Source: New Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 150 of 154 29 February 2012 Associated objective(s):

SR-0001-K13 (Freq: LIC=)

Describe the relationship between QGAs, Station Operating Procedures (QCOA, QCOP, and QCGP) and Severe Accident Management Guidelines (SAMG).

2.4.16 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. (RO=3.5 / SRO=4.4)

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 151 of 154 29 February 2012 74 ID: QDC.ILT.15653 Points: 1.00 (A reference is provided for this question)

Given the following conditions:

Unit 1 is at rated power Unit 2 is in Cold Shutdown preparing for a refueling outage It has been determined that QCARP entry IS required Which one of the following QCARPs will be used for a fire in the Safe Shutdown Makeup Pump Room?

A.

0030-01 B.

0030-02 C.

0030-03 D.

0030-04 Answer:

C Answer Explanation:

QCOA 0010-12 directs the QCARPs to be entered if QGAs will be ineffective in maintaining the Unit in Hot Shutdown or reaching a Cold Shutdown. Unit 2 is already in a Cold Shutdown condition and will not enter a QCARP. QCOA 0010-12 attachment C, defines the designated area for U1 Safe Shutdown Makeup Pump fire as TB-II. QCOA 0010-12 attachment D lists QCARP 0030-03 as the procedure to use for a fire in TB-II.

Distractor 1 is incorrect: Plausible because QCARP 0030-01 is used for fires in TB-III, which is directly adjacent to TB-II.

Distractor 2 is incorrect: Plausible because QCARP 0030-02 is used for fires in TB-I, which is directly adjacent to TB-II.

Distractor 3 is incorrect: Plausible because QCARP 0030-04 is used for fires in TB-II as well, but for Unit 2.

Reference:

QCOA 0010-12 rev 37 Reference provided during examination: QCOA 0010-12 rev 37 Cognitive level: High Level (RO/SRO): RO Tier: 3 K/A: 2.4 EMERGENCY PROCEDURES / PLAN 2.4.27 Knowledge of fire in the plant procedures (RO=3.4 / SRO=3.9) 10 CFR Part 55 Content: 41.10 Question Source: New Question History: N/A Comments:

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 152 of 154 29 February 2012 Associated objective(s):

2.4.27 Knowledge of fire in the plant procedures (RO=3.4 / SRO=3.9)

SR-4100-P20 (Freq: LIC=B)

Given an operating reactor plant when a severe uncontrolled fire which renders the appropriate normal and abnormal procedures inadequate occurs, perform the control room actions in accordance with QCOA 0010-12 and approriate QCARP procedure.

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 153 of 154 29 February 2012 75 ID: QDC.ILT.15655 Points: 1.00 Given:

Unit 1 is operating at rated power.

1A RPS is supplied from normal power.

1B RPS is on alternate power.

A fire starts in the cabling under the Center Desk area resulting in heavy smoke and toxic fumes rapidly spreading throughout ALL of the areas supplied by the Control Room HVAC.

The Shift Manager orders evacuation of the control room due an uninhabitable control room envelope.

Unit 1 was UNABLE to be scrammed prior to abandoning the control room.

The Shift Manager directs the U1ANSO to scram the reactor.

Under these circumstances, how will Unit 1 reactor be scrammed?

The U1ANSO will...

A.

Open the 1A RPS feed at MCC 18-2 and the 1B RPS feed at MCC 15-2.

B.

Open the 1A RPS feed at MCC 15-2 and the 1B RPS feed at MCC 19-2.

C.

Open the breakers that feed the RPS 901-15 & 17 panels in the Auxiliary Electric Room.

D.

Manually trip the 1A & 1B RPS Motor Generator Sets drive motors from the Auxiliary Electric Room.

Answer:

A Answer Explanation:

The Auxiliary Electric Room part of the control room envelope and is unavailable due to the heavy smoke and toxic fumes. Therefore, with RPS 1A on normal feed the power supply breaker must be opened at 18-2 and with RPS 1B on reserve feed the reserve supply breaker on 15-2 needs to be opened.

Distractor 1: This will not scram the reactor with the stated power supply lineup Distractor 2: The Aux Elec Room is inaccessible Distractor 3: The Aux Elec Room is inaccessible

Reference:

QOA 0010-05 rev 24 Reference provided during examination: N/A Cognitive level: High Level (RO/SRO): RO Tier: 3 Group: N/A

EXAMINATION ANSWER KEY U.S. Nuclear Regulatory Commission 2012 RO Written Exam (Quad Cities)

OPS MASTER STANDALONE Page: 154 of 154 29 February 2012 K/A: 2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. (RO=4.2 / SRO=4.1) 10 CFR Part 55 Content: 41.10 Question Source: Quad Cities ILT Exam Bank Question History: N/A Comments:

Associated objective(s):

2.4.34 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects. (RO=4.2 / SRO=4.1)

SRN-EVAC-K09 (Freq: LIC=B NF=B)

Given a Control Room Evacuation, STATE the locations to which your job position may be assigned.