ML11206A563

2011 Duane Arnold Energy Center Initial License Examination Administered Reactor Operator Written Exam

ML11206A563
Person / Time
Site:	Duane Arnold
Issue date:	07/08/2011
From:	Chuck Zoia Operations Branch III
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Download: ML11206A563 (150)

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Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 206000 K1.10 Importance Rating 3.4 Knowledge of the physical connections and/or cause- effect relationships between HIGH PRESSURE COOLANT INJECTION SYSTEM and the following: Condensate storage and transfer system: BWR-2,3,4 Question: RO Question # 1 While HPCI is in a CST to CST lineup for surveillance testing the following occur:

• Annunciator 1C03C (D-4) TORUS HI LEVEL HPCI SUCTION TRANSFER INITIATE alarms.

• MO-2321 INBD TORUS SUCTION ISOLATION and MO-2322 OUTBD TORUS SUCTION ISOLATION open.

Which one of the following is the correct system response?

MO-2300 CST SUCTION closes when __(1)__

When either MO-2321 or MO-2322 is full open, __(2)___ will automatically close.

A. (1) both MO-2321 and MO-2322 are full open (2) only CV-2315 TEST BYPASS B. (1) either MO-2321 or MO-2322 are full open (2) only CV-2315 TEST BYPASS C. (1) both MO-2321 and MO-2322 are full open (2) both CV-2315 TEST BYPASS and MO-2316 REDUNDANT SHUTOFF D. (1) either MO-2321 or MO-2322 are full open (2) both CV-2315 TEST BYPASS and MO-2316 REDUNDANT SHUTOFF Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - [part 1 correct, part 2 incorrect] MO-2300, CST SUCTION valve automatically closes when both MO-2321 AND MO-2322 INBD and OUTBD TORUS SUCTION ISOLATION Valves reach full open.

When MO-2321 full open or MO-2322 full open CV-2315, Test Bypass AND MO-2316, Redundant Shutoff Valves close.

B. Incorrect - [part 1 incorrect, part 2 incorrect] MO-2300, CST SUCTION valve automatically closes when both MO-2321 AND MO-2322 INBD and OUTBD TORUS SUCTION ISOLATION Valves reach full open.

When MO-2321 full open or MO-2322 full open CV-2315, Test Bypass AND MO-2316, Redundant Shutoff Valves close.

C. Correct - MO-2300 CST SUCTION valve automatically closes when both MO-2321 AND MO-2322 INBD and OUTBD TORUS SUCTION ISOLATION Valves reach full open.

When MO-2321 reaches full open or MO-2322 reaches full open, CV-2315 Test Bypass AND MO-2316 Redundant Shutoff valves close.

D. Incorrect - [part 1 incorrect, part 2 incorrect] MO-2300, CST SUCTION valve automatically closes when both MO-2321 AND MO-2322 INBD and OUTBD TORUS SUCTION ISOLATION Valves reach full open.

When MO-2321 full open or MO-2322 full open CV-2315, Test Bypass AND MO-2316, Redundant Shutoff Valves close.

OI-152, Section 8.1 & 8.2, pgs 31

& 32.

Technical Reference(s): (Attach if not previously provided)

SD 152, pg 29 SD 537 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

DAEC RO Bank Question Source: Bank #

19199 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-9 NRC review SAT 6/14/11 - found issues with question (lack of only/both) rearranged part b slightly 6/15 - changed to either/or for a2/c2, clarified explanations ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 211000 K1.01 Importance Rating 3.0 Knowledge of the physical connections and/or cause- effect relationships between STANDBY LIQUID CONTROL SYSTEM and the following: Core spray line break detection: Plant-Specific Question: RO Question # 2 Which one of the following describes the relationship between the Standby Liquid Control System (SBLC) and the Core Spray (CS) line break detection system?

A differential pressure switch measures the pressure difference between the ____(1)_____

AND the inside of the ______(2)_______

A. (1) below core plate (inner pipe of the SBLC penetration)

(2) reactor pressure vessel in the downcomer annulus region.

B. (1) above core plate (outer pipe of the SBLC penetration)

(2) reactor pressure vessel in the downcomer annulus region.

C. (1) below core plate (inner pipe of the SBLC penetration)

(2) CS sparger pipe, just outside the reactor vessel.

D. (1) above core plate (outer pipe of the SBLC penetration)

(2) CS sparger pipe, just outside the reactor vessel.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - A differential pressure switch measures the pressure difference between the bottom of the core which is the outer pipe of the SBLC penetration. The inside of the core spray sparger pipe measures the pressure inside the core shroud.

B. Incorrect - The inside of the core spray sparger pipe measures the pressure inside the core basket.

C. Incorrect - A differential pressure switch measures the pressure difference between the bottom of the core which is the outer pipe of the SBLC penetration.

D. Correct - A differential pressure switch measures the pressure difference between the bottom of the core which is the outer pipe of the SBLC penetration. The inside of the core spray sparger pipe measures the pressure inside the core shroud.

Technical Reference(s): SD 151, pgs 20 - 22 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 400000 K2.02 Importance Rating 2.9 Knowledge of electrical power supplies to the following: CCW valves Question: RO Question # 3 The plant is operating in MODE 1 at 100% power with the following conditions:

• The Startup Transformer is removed from service due to preplanned maintenance

• The Standby Transformer is powering busses 1A3 and 1A4

• A LOCA occurs

• RPV level lowered to 30 inches before recovering to 175 inches What is the response to this event, if any, of the RBCCW Drywell Supply and Return Isolation Valves, MO-4841A and MO-4841B?

MO-4841A and MO-4841B will ________.

A. remain OPEN.

B. go closed and cannot be reopened due to a loss of power to the valves.

C. go closed but can be manually re-opened with no additional operator action.

D. go closed and will require operator actions to reset the isolation and open the valves.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Valves will close on Group 7 isolation B. Incorrect - Valves are powered by 1B42 (essential power)

C. Incorrect - Group 7 required to be reset on 1C31 prior to opening valves.

D. Correct - The valve solenoids for the Drywell Cooling Isolation Valves (CV) are powered by 120 VAC Instrument AC from 1Y11 and 1Y21, and are Energize-to-Close. The Motor-Operated valves for RBCCW are powered from 480 VAC 1B42.

For Group 7, the RBCCW and Well Water Isolations Seal In with the use of the Aux Relay, CR-4841X. When the Reactor Low-Low-Low Level Sensor Relays reset, the Reset pushbutton on 1C31 will need to be depressed to reset the Group 7 Isolation signal. There is an amber indicating light at 1C31 to indicate when the Isolation Signal is Locked In. When the Isolation Signal is Reset, then the Drywell Cooling solenoid valves will reopen automatically if Drywell Cooling is on, but the motor-operated valves for RBCCW will need to be reopened.

SD 414, pg 10 Technical Reference(s): (Attach if not previously provided)

SD 959-1, pg 40, 43 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

6-3-11-reworded distractors 6-9 NRC OK enhancement ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 300000 K2.02 Importance Rating 3.0 Knowledge of electrical power supplies to the following: Emergency air compressor Question: RO Question # 4 The plant conditions are as follows:

• Backup Instrument Air Compressor 1K1 is in the STANDBY-operating mode

• 1K1 electrical power is being supplied from 480 VAC Bus 1B33 A large electrical disturbance occurs resulting in:

• LLRPSF transformers XR1 and XR2 de-energizing, and

• A Bus 1A3 lockout.

Which one of the following describes the response of the Backup Instrument Air Compressor 1K1?

1K1 will ______.

A. need to have its power supply transferred from 1B33 to 1B45 to start B. start when header pressure reaches 100 psig and will cycle to maintain 100 - 110 psig C. start when header pressure reaches 90 psig and will cycle to maintain 90 - 100 psig D. need to have HSS-3002, BACKUP COMPRESSOR 1K-1 PRESSURE SELECT SWITCH placed in the PRIMARY position to start Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - The transfer switch 1N3312 is a "break before make" type which is operated to allow 1K-1 to be powered from either 1B3312 or 1B4501. 1B3312 will be the normal power supply selected. If 1B33[1B45] has to be de-energized for any reason, the compressor power can be transferred to 1B4501[1B3312].

B. Incorrect - This condition is not a trip, but without power the compressor does not run.

C. Incorrect - This condition is not a trip, but without power the compressor does not run.

D. Incorrect - This condition is not a trip, but without power the compressor does not run.

Technical Reference(s): OI-518.1, Sect 4.7, pg 27 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # DAEC RO 19111 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215005 K3.08 Importance Rating 3.0 Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following: core thermal calculations Question: RO Question # 5 The plant is operating in MODE 1 at 93% power with the following plant conditions:

• A and B APRMs are bypassed to support LPRM whisker burns.

• LPRM 4D-08-09, an LPRM shared between APRM A and B fails upscale.

Which of the following describes the affect of this failure on the value of computer point C179, NSSS1 CORE THERMAL POWER (MWTH)?

A. "B" APRM reading will increase causing C179 to RISE B. "B" APRM reading will increase, however, since the APRM is bypassed C179 will REMAIN THE SAME C. LPRMs do NOT input into the Reactor Heat Balance Equation and therefore C179 will REMAIN THE SAME D. "B" APRM readings will lower because the "D" Level LPRM upscale reading is automatically rejected causing C179 to LOWER Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The affect on the B APRM is correct but it has no effect on the heat balance B. Incorrect - The affect on the B APRM is correct but it has no effect on the heat balance C. Correct - The heat balance is used to adjust APRM gains, LPRMs and APRMs are not inputs to MWTH D. Incorrect - LPRMs are not automatically rejected in APRMs, however in the RBM system they are.

SD-878.3, Rev 8; Pages 44-45.

Technical Reference(s): SD-900, Rev. 4, pgs. 7-9. (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 81.01.01.15 Learning Objective: (As available)

Question Source: Bank # 2005 NRC Exam Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2005 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2 55.43 General Design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 212000 K3.10 Importance Rating 3.5 Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION SYSTEM will have on following: The ability of the core cooling systems to provide adequate core cooling during loss of coolant accidents Question: RO Question # 6 Plant conditions are as follows:

• Manual and automatic actions have failed to insert control rods

• RPV Flooding EOP has been entered How is adequate core cooling assured during this event?

Depressurize the RPV, then control injection to establish and maintain ___(1)___. The core will then be cooled by ___(2)___.

A. (1) RPV level between -25 in. and +211 (2) submergence or Steam Cooling B. (1) RPV level between -25 in. and the level required to lower power below 5%

(2) full submergence C. (1) RPV pressure above the Minimum Steam Cooling Pressure (2) submergence or Steam Cooling D. (1) RPV level flooded to the elevation of the RPV flange and RPV pressure a minimum of 150 psig above Torus pressure (2) Steam Cooling Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - This is the broad range of water level requirements during an ATWS it would not apply if RPV Flooding is entered.

B. Incorrect - This is the broad range of water level requirements during an ATWS it would not apply if RPV Flooding is entered.

C. Correct - RPV flooding, is used to cool the core when RPV water level cannot be determined. The specified actions first depressurize the RPV, then control injection to establish and maintain one of the following conditions:

• The RPV flooded to the elevation of the main steam lines. The core will then be cooled by full submergence. This condition may ultimately be achieved under either shutdown or failure-to-scram conditions.

• RPV pressure above the Minimum Steam Cooling Pressure. The core will then be cooled by submergence or steam cooling. Since reactor power must be at least 6%-

10% to generate the amount of steam required to sustain the Minimum Steam Cooling Pressure, this condition is applicable only under ATWS conditions.

D. Incorrect - The direction of RPV/F EOP is to maintain water level at the Main Steam Lines, not the RPV head. The 150 psig is the minimum steam cooling pressure for 4 SRVs open.

EOP RPV Flooding Bases pg 2 Technical Reference(s): RPV Flooding step F-7 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 217000 K4.04 Importance Rating 3.0 Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design feature(s) and/or interlocks which provide for the following: Prevents turbine damage: Plant-Specific Question: RO Question # 7 During a manual start of RCIC, the following indications are observed:

• TURBINE STEAM SUPPLY MO-2404 starts to open

• RCIC Turbine speed begins to rise

• RCIC Pump Discharge pressure begins to rise At this point annunciator 1C04C, A-5, RCIC MO-2405 TURB TRIP alarms followed 5 seconds later, by the following alarms:

• Annunciator 1C04C, D-9, RCIC TURBINE BEARING OIL LO PRESSURE alarms.

• Annunciator 1C04C, B-4, RCIC LO FLOW alarms.

• Reactor water level is 186 inches and stable.

No other alarms are present on 1C04C and all alarms are in proper working order.

Which one of the following provides the correct analysis of this situation?

A. A turbine trip has occurred on low flow.

B. A turbine trip has occurred on overspeed.

C. A turbine trip has occurred on low oil pressure.

D. A turbine trip has occurred on low pump suction pressure.

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - A low flow condition would not cause a turbine trip B. Correct - A turbine overspeed trip will only cause an alarm on the turbine trip. The low oil pressure and low flow result from the turbine speed coasting down after the RCIC turbine trip.

C. Incorrect - During a loss of oil pressure the turbine will overspeed because the RCIC turbine control valve is opened by spring pressure and closed by oil pressure.

D. Incorrect - There is no indication that MO-2404 failed to close.

Technical Reference(s): OI-150, pg 48 and SD -150 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-9-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 205000 K4.03 Importance Rating 3.8 Knowledge of SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) design feature(s) and/or interlocks which provide for the following: Low reactor water level:

Plant-Specific Question: RO Question # 8 The plant is operating in MODE 4 in Shutdown Cooling with the following conditions:

• RPV level is 190 inches

• The A RHR pump and A RHRSW pump are running An event occurs that causes RPV level to rapidly drop to 50 inches.

Which one of the following describes how the RHR pumps automatically respond to the signal?

A RHR Pump C RHR Pump A. Trips and restarts Starts and operates when the system on minimum flow automatically realigns to a LPCI mode B. Trips and does not Starts and operates restart on minimum flow C. Trips and does not Attempts to start and restart immediately trips D. Trips and restarts Attempts to start and when the system immediately trips automatically realigns to a LPCI mode Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The "A" pump trips. All others start but C trips. the RHR System will not automatically realign itself for LPCI injection B. Incorrect - The "A" pump trips and does not restart. All others start but C trips C. Correct - Per SD 149, page 22 - In the event a LOCA occurs when the RHR System is in the shutdown cooling mode, the RHR System will not automatically realign itself for LPCI injection. Operator actions required to initiate the LPCI mode of RHR include resetting the Group 4 Isolation Seal-In, restoring torus suction flowpath to the RHR pumps, and manually restarting the RHR pumps that have tripped.

Additionally, the SDC suction valves close on the LPCI signal (PCIS Group 4). The "C" RHR Pump breaker will receive a start signal but immediately trip. The trip occurs due to no suction path present to prevent pump damage. This is NOT a start permissive, it s a pump trip (SD-149 page 12)

D. Incorrect - The "A" pump trips. The "C" RHR Pump breaker will receive a start signal but immediately trip. The trip occurs due to no suction path present to prevent pump damage. This is NOT a start permissive, it s a pump trip (SD-149 page 12). The RHR System will not automatically realign itself for LPCI injection SD 149 Rev 11 pages 12 & 22 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 2005 NRC Exam Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2005 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3 revised a RHR pump distractors A and D 6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 218000 K5.01 Importance Rating 3.8 Knowledge of the operational implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION SYSTEM: ADS logic operation Question: RO Question # 9 The plant was operating in MODE 2 at 7% power when an accident occurred. Current plant conditions are as follows:

• DW pressure is 8 psig, rising

• HPCI system tripped

• RCIC injecting into RPV @ 415 gpm

• RPV level reaches 64 inches and lowering at Time Zero (T0)

Assuming no operator action, the ADS system will automatically actuate to lower RPV pressure when any lower pressure ECCS pump ___(1)___ with ___(2)___ (referenced to time zero).

A. (1) breaker is CLOSED (2) a 5 second time delay B. (1) breaker is CLOSED (2) a 2 minute time delay C. (1) reaches normal discharge pressure (2) a 5 second time delay D. (1) reaches normal discharge pressure (2) a 2 minute time delay Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Breaker closed is not the correct signal; triple low level must be in place two minutes. 5 seconds associated with CS pump B. Incorrect - Breaker closed is not the correct signal C. Incorrect - ADS waits two minutes. 5 seconds associated with CS pump D. Correct - Timer starts when reactor water level reaches low-low-low level. Two minutes later, if an RHR or Core Spray pump is at normal discharge pressure, ADS will open 4 SRVs. This assumes that timers are not overridden SD-183.1 Rev. 6, Page 17 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 8.02.01.02 Learning Objective: (As available)

Question Source: Bank # 2005 NRC Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2005 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 8 55.43 Components, capacity, and functions of emergency systems.

Comments:

6-3 changed to 90 seconds from no time delay 6-9-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 203000 K5.02 Importance Rating 3.5 Knowledge of the operational implications of the following concepts as they apply to RHR/LPCI: INJECTION MODE (PLANT SPECIFIC): Core cooling methods Question: RO Question # 10 A plant shutdown is in progress and conditions are as follows:

• Reactor is in MODE 3 with RPV pressure 30 psig

• Reactor water level is 190 inches on all GEMAC level instruments

• Both reactor recirculation pumps are shutdown

• GSW is shutdown and being drained for maintenance

• A loss of Shutdown Cooling occurs and RHR CANNOT be recovered Which one of the following would provide an alternate method to ensure core DECAY HEAT REMOVAL is re-established?

(Assume no Defeats are installed.)

A. Start RCIC in CST-To-CST mode to lower RPV pressure B. Raise RPV level to +214 inches using HPCI to provide natural circulation.

C. Starting one of the Reactor Recirculation pumps to re-establish recirculation flow.

D. Raise RPV level with the A Core Spray pump and perform feed and bleed to the torus with SRVs.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Reactor pressure is below RCIC isolation setpoint making RCIC unavailable.

B. Incorrect - Reactor pressure is below HPCI isolation setpoint making HPCI unavailable.

C. Incorrect - Not a required action IAW AOP and with GSW OOS, cannot be started by procedure.

D. Correct - Because these actions are consistent with guidance in AOP-149, Inadequate Decay Heat Removal.

AOP-149, Sect 4.2, pg 7 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTSI 11421 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-3-1-changed distractor C 6-9 changed stem, removed distractor revision. NRC OK 6/14/11 - changed distractor D to A Core Spray pump, condensate pump cooled by GSW.

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 209001 K6.11 Importance Rating 3.6 Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM: ADS Question: RO Question # 11 The plant was operating in MODE 1 at 44% power with the following conditions:

• HPCI was inoperable for preplanned maintenance A LOCA then occurred resulting in the following plant conditions:

• DW Pressure is 7 psig rising slowly

• All control rods fully inserted

• RPV Pressure is 730 psig, lowering slowly

• RPV Level is 60 inches, lowering slowly

• ADS timers initiated and are timing out

• With 30 seconds left on the ADS timers, the A ADS timer loses power Which one of the following describes the status the ADS Valves and Core Spray Pump(s) when the B ADS logic times out?

ADS Valves ___(1)___

Core Spray Pump(s) _____(2)_____

A. (1) PSVs will remain closed.

(2) B ONLY remains on minimum flow.

B. (1) PSV 4400 and PSV 4405 only will open.

(2) "A" and "B" inject when pressure lowers below their discharge head.

C. (1) PSV 4400, PSV 4402, PSV 4405 and PSV 4406 will open.

(2) "B" ONLY injects when pressure lowers below its discharge head.

D. (1) PSV 4400, PSV 4402, PSV 4405 and PSV 4406 will open.

(2) "A" and "B" inject when pressure lowers below their discharge head.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Although the channel A timer is not energized, ADS will actuate with only one channel timed out.

B. Incorrect - Only one timer needs to time out to actuate the all ADS valves.

C. Incorrect - The loss of power to the ADS logic does not affect the Core Spray pumps since this logic is not shared, both pumps will inject.

D. Correct Only one timer needs to time out to actuate the all ADS valves and the ADS logic does not affect the Core Spray pumps, both pumps will inject.

Technical Reference(s): SD-183.1, pg 15 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3-11 added PSV to answer/distractors 6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 261000 K6.03 Importance Rating 3.0 Knowledge of the effect that a loss or malfunction of the following will have on the STANDBY GAS TREATMENT SYSTEM : Emergency diesel generator system Question: RO Question # 12 The plant is operating in MODE 1 at 100% power with the following plant conditions:

• The B SBDG is tagged out for heat exchanger replacement.

• A tornado strikes the switchyard causing a loss of off-site power (LOOP).

Assuming no operator action, which one of the following is the status of the Standby Gas Treatment (SBGT) systems?

A. Both SBGT trains remain in STANDBY and are available to start on an initiation signal B. ONLY the "A" SBGT has received a start signal and it has automatically started C. Both SBGT lockout relays tripped but only the A SBGT train is running D. Both SBGT lockout relays have tripped and both SBGT trains are running Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The loss of off-site power results in a loss of RPS which causes an initiation of both SBGT systems. The 480V Bus 1B34 will be supplied by the "A" Diesel which will allow an auto start of SBGT "A".

B. Incorrect - The loss of off-site power results in a loss of RPS which causes an initiation of both SBGT systems. The 480V Bus 1B34 will be supplied by the "A" Diesel which will allow an auto start of SBGT "A".

C. Correct - The loss of off-site power results in a loss of RPS which causes an initiation of both SBGT systems. The 480V Bus 1B34 will be supplied by the "A" Diesel which will allow an auto start of SBGT "A".

D. Incorrect - The loss of off-site power results in a loss of RPS which causes an initiation of both SBGT systems. The 480V Bus 1B34 will be supplied by the "A" Diesel which will allow an auto start of SBGT "A".

Technical Reference(s): AOP-358, ARP-1C05B (C-8) (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215003 A1.02 Importance Rating 3.7 Ability to predict and/or monitor changes in parameters associated with operating the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM controls including: Reactor power indication response to rod position changes Question: RO Question # 13 A reactor startup from a 6 day maintenance outage is in progress. The reactor is in MODE 2 and control rod withdrawal is in progress with power in the IRM range.

As power rises, the IRM range switches shall be moved to maintain the IRM indication between

___(1)___ on the odd scale and between ___(2)___ on the even scale.

A. (1) 3/40 and 25/40 (2) 10/125 and 75/125 B. (1) 10/125 and 75/125 (2) 3/40 and 25/40 C. (1) 10/40 and 25/40 (2) 25/125 and 100/125 D. (1) 25/125 and 100/125 (2) 10/40 and 25/40 Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - IAW OI-878.2, Continue to reposition the IRM range switches to maintain indications on the IRM recorders between 10/125 and 75/125 on the Even scale and between 3/40 and 25/40 on the Odd scale.

B. Incorrect - Indication should be between 10/125 and 75/125 on the Even scale and between 3/40 and 25/40 on the Odd scale.

C. Incorrect - Indication should be between 10/125 and 75/125 on the Even scale and between 3/40 and 25/40 on the Odd scale.

D. Incorrect - Indication should be between 10/125 and 75/125 on the Even scale and between 3/40 and 25/40 on the Odd scale.

Technical Reference(s): OI-878.2, pg 7 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3 changed to 2 part answer by moving odd and even to stem 6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 223002 A1.03 Importance Rating 2.5 Ability to predict and/or monitor changes in parameters associated with operating the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF controls including: SPDS/ERIS/CRIDS/GDS: Plant-Specific Question: RO Question # 14 The plant is operating in MODE 1 at 100% power with the following plant conditions:

• The B RPS MG set is to be secured to support planned maintenance

• The RPS Half Scram Preparation checklist is in progress

• The CRS directs that Reactor Water Cleanup be secured Which one of the following actions may need to be performed in accordance with OI 261, Reactor Water Cleanup System for the above conditions?

A. Substitute RWCU System Flow computer point (B017) to indicate zero to maintain an accurate heat balance.

B. Open MO-2732, "RWCU Drain to Radwaste", to ensure the system depressurizes completely while it is isolated.

C. Inform Chemistry that the RWCU system is isolated and to commence taking manual RWCU system grab samples.

D. Isolate the Non-Regenerative Heat Exchanger by isolating the shell side RBCCW flow before isolating the tube side RWCU flow.

Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - IAW OI-261, Computer point B017, RWCU System Flow, may need to be substituted to zero, during system shutdown/isolation, to maintain accurate 3D Monicore periodic logs.

B. Incorrect - There is no need to drain the system.

C. Incorrect - Manual grab samples would be required if the system was operating and the normal sampling system was not operable.

D. Incorrect - The entire system is to be isolated not the Non-Regenerative Heat Exchanger.

Technical Reference(s): OI-261, pg 4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # Sys ID 18933 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-3-11 revised stem although dont like using may.

6-9-11-NRC OK - enhancement ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 239002 A2.01 Importance Rating 3.0 Ability to (a) predict the impacts of the following on the RELIEF/SAFETY VALVES ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Stuck open vacuum breakers Question: RO Question # 15 The plant is operating at 100% power with the following conditions:

• A spurious Group 1 isolation occurs

• Low Low Set (LLS) SRVs actuate to control pressure

• One LLS SRV tailpipe vacuum breaker is stuck open such that Containment pressure is 1.2 psig and rising slowly (1) What is the result of this condition? AND (2) What actions need to be taken?

A. (1) Steam from the SRV will go into the Drywell atmosphere (2) Install EOP Defeat 9 and vent the drywell via SBGT.

B. (1) Steam from the SRV will go into the Drywell atmosphere (2) AOP 573 may be used to vent the drywell via SBGT as long as containment pressure is < 2.0 psig.

C. (1) Steam from the SRV will go into the Torus atmosphere (2) Install EOP Defeat 9 and vent the drywell via SBGT.

D. (1) Steam from the SRV will go into the Torus atmosphere (2). AOP 573 may be used to vent the drywell via SBGT as long as containment pressure is < 2.0 psig.

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The SRV vacuum breaker being open allows direct communication of some steam to the DW air space NOT the Torus airspace. Defeat 9, High Drywell Pressure and RPV low level defeat is not authorized in this situation.

B. Correct - The SRV vacuum breaker being open allows direct communication of some steam to the DW air space that may raise DW pressure. AOP-573 directs venting the DW if pressure rises to 1.0 to 1.5 psig by venting Drywell through SBGT.

C. Incorrect - The SRV vacuum breaker being open allows direct communication of some steam to the DW air space NOT the Torus airspace. Defeat 9, High Drywell Pressure and RPV low level defeat is not authorized in this situation.

D. Incorrect - The SRV vacuum breaker being open allows direct communication of some steam to the DW air space that may raise DW pressure. AOP-573 directs venting the DW if pressure rises to 1.0 to 1.5 psig by venting Drywell through SBGT.

AOP-573 Technical Reference(s): SD 183-1, pg 19 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 259002 A2.05 Importance Rating 3.2 Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of applicable plant air systems Question: RO Question # 16 The plant is starting up in Mode 1 at 12% power with the following conditions:

• A RFP is in service

• The Startup Feedwater Control Valve CV-1622 is in service in Auto Which one of the following describes how a loss of Instrument Air will affect CV-1622 and what actions are required to control Reactor water level?

Feedwater Startup Control Valve CV-1622 fails ___(1)___ .

Control Reactor water level by ___(2)___ IAW AOP 644, FEEDWATER/ CONDENSATE MALFUNCTION.

A. (1) open (2) throttling the Startup Feedline Block Valve MO-1631 CLOSED B. (1) closed (2) OPENING Feed Regulating Valve CV-1579 as appropriate C. (1) locked up (as-is)

(2) tripping feedwater pumps or throttling Feed Regulating Valve CV-1579 as appropriate.

D. (1) locked up (as-is)

(2) throttling the Startup Feedline Block Valve MO-1631 Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - With a loss of air the Startup Feed Reg Valve will lock up (fail as-is). If the failure lasts longer than 30 minutes, the FRV will tend to drift open (even locked up).

B. Incorrect - With a loss of air the Startup Feed Reg Valve will lock up (fail as-is). If the failure lasts longer than 30 minutes, the FRV will tend to drift open (even locked up).

C. Incorrect - With a loss of air the Startup Feed Reg Valve will lock up (fail as-is). If the failure lasts longer than 30 minutes, the FRV will tend to drift open (even locked up).

There is no direction to trip the feedwater pumps to maintain Reactor water level.

D. Correct - With a loss of air the Startup Feed Reg Valve will lock up (fail as-is). If the failure lasts longer than 30 minutes, the FRV will tend to drift open (even locked up).

ARP-1C05A, E-1 directs throttling Blocking Valve MO-1631 or opening Feed Reg Valve CV-1579(1621) as appropriate.

ARP-1C05A, E-1 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3-11 removed AOP as reference. In distractor C changed closed to throttling.

6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 264000 A3.06 Importance Rating 3.1 Ability to monitor automatic operations of the EMERGENCY GENERATORS (DIESEL/JET) including: Cooling water system operation Question: RO Question # 17 The plant is starting up in MODE 3 with the following conditions:

• Reactor Pressure at 675 psig

• Both ESW pumps were operating to support torus cooling operations.

• A loss of offsite power (LOOP) occurs with all systems operating as designed.

Which one of the following correctly states:

(1) When will the ESW pumps restart?

(2) What is the ESW flowrate compared to prior to the loss of offsite power (more or less)?

A. (1) when the SBDGs are supplying the bus (2) less B. (1) when the SBDGs are supplying the bus (2) more C. (1) 2 minutes after the SBDGs are supplying the bus (2) less D. (1) 2 minutes after the SBDGs are supplying the bus (2) more Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - ESW flow will be greater than before the LOOP because the cooling water valves for the SBDG will open under control of the SBDG start logic.

B. Correct - The ESW pumps start automatically if the associated emergency diesel generator starts. ESW flow will be greater than before the LOOP because the cooling water valves for the SBDG will open under control of the SBDG start logic.

C. Incorrect - The ESW pumps start automatically if the associated emergency diesel generator starts. ESW flow will be greater than before the LOOP because the cooling water valves for the SBDG will open under control of the SBDG start logic.

D. Incorrect - The ESW pumps start automatically if the associated emergency diesel generator starts.

Technical Reference(s): SD-454, pg 7 & 8. (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

6-9 NRC OK - explanations fixed ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 262002 A3.01 Importance Rating 2.8 Ability to monitor automatic operations of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) including: Transfer from preferred to alternate source Question: RO Question # 18 The plant is operating in MODE 1 at 100% power with the following conditions:

• The 1Y23 Power Source Manual Transfer Switch (HSS-1Y23A) is in the AUTO TO 1Y2 position

• The voltage at 1Y23 lowers to 100 VAC and then recovers to 120 VAC Which ONE of the following describes the affect of this transient on Uninterruptible Power System loads?

Loads will be A. continuously powered from 1D45/1Y4.

B. interrupted by a momentary BREAK BEFORE MAKE transfer to 1Y2 and remain powered from 1Y2.

C. continuously powered during the MAKE BEFORE BREAK transfer to 1Y2 and then automatically transfer back to 1D45/1Y4 when voltage recovers.

D. interrupted by a momentary BREAK BEFORE MAKE transfer to 1Y2 and then automatically transfer back to 1D45/1Y4 when voltage recovers.

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - This would be true if voltage lowered to 115 VAC and recovered.

B. Correct - When voltage lowers to 105 VAC, device 27-22 forces a break before make transfer to 1Y2. Operator action is required to enable transfer back to 1D45/1Y4.

C. Incorrect - This would be true if 1Y22 operated like the Static Switch.

D. Incorrect - This would be true if 1Y23 Power Source Manual Transfer Switch (HSS-1Y23A) were in the 1D45/1Y4 position.

Technical Reference(s): SD-357 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 2007 NRC Exam Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 262001 A4.04 Importance Rating 3.6 Ability to manually operate and/or monitor in the control room: Synchronizing and paralleling of different A.C. supplies Question: RO Question # 19 The plant was operating in MODE 1 at 100% power with the following conditions:

• A severe electrical transient has occurred resulting in a station blackout

• AOP 301.1, Station Blackout, has been entered

• The grid operator reports that power has been restored to the DAEC switchyard

• Normal voltage conditions are expected to be restored within the next 30 minutes The BOP reports the following from 1C08:

• The GENERATOR OUTPUT H BREAKER Synchronizing Switch is ON

• The RUNNING voltmeter reads 82 volts Can the Essential Buses 1A3 and 1A4 be restored using the Standby Transformer until normal voltage is restored to the grid?

A. No, because the Degraded Voltage Relays cannot be reset with the Synchronizing Switch ON B. Yes, provided the Degraded Voltage Relays are reset at 1C08 only. An override at 1C351/1C352 is not required C. No, because the Degraded Voltage Relays cannot be reset at 1C08 OR overridden at 1C351/1C352.

D. Yes, provided the Degraded Voltage Relays are be reset at 1C08 and then overridden at 1C351/1C352.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The low voltage can be overridden. Sych switch position has no impact B. Incorrect - The degraded voltage can NOT be reset at this voltage, voltage must be above 96% (111 volts) to reset.

C. Incorrect - The low voltage can be overridden.

D. Correct - Overriding the degraded voltage will work if incoming voltage is more than 65% (2700 Volts) (incoming of 78 volts). If degraded grid voltages exist, override degraded bus voltage condition on essential buses 1A3/1A4 by resetting the degraded voltage relays at 1C08 by pushing the degraded voltage reset pushbuttons, then override the Degraded Voltage Relays at 1C351[1C352] using TEST switches.

Technical Reference(s): AOP-301.1, pg 19 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # DAEC Bank #19551 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-9 NRC OK with change - enhancement ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215004 A4.06 Importance Rating 3.2 Ability to manually operate and/or monitor in the control room: Alarms and lights Question: RO Question # 20 The reactor is in MODE 2 with a reactor startup in progress with the following conditions:

• No SRMs or IRMs are bypassed

• The SRM detectors are being withdrawn per IPOI-2, Startup Which one of the following sets of conditions will result in activation of alarm 1C05A (E-5), SRM DETECTOR RETRACTED WHEN NOT PERMITTED?

All IRM Range A SRM Reading B SRM Reading C SRM Reading D SRM Reading Switch Positions A. 1 120 cps 120 cps 120 cps 120 cps B. 2 90 cps 150 cps 150 cps 150 cps C. 3 90 cps 120 cps 150 cps 120 cps D. 4 90 cps 120 cps 120 cps 120 cps Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - plausible; would be true if SRM counts were given below 100 cps B. Correct - With detectors partially withdrawn, an SRM reading 90 cps will generate SRM DETECTOR RETRACTED WHEN NOT PERMITTED alarm with IRMs on range 2.

C. Incorrect - plausible; would be true if IRMs were given below range 3 D. Incorrect - plausible; would be true if IRMs were given below range 3 Technical Reference(s): ARP 1C05A E-5 Rev 58 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTSI 11263 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

6-3-11 revised distractor c numbers.

6-9 Revised C - NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 263000 2.2.22 Importance Rating 4.0 Equipment Control: Knowledge of limiting conditions for operations and safety limits.

Question: RO Question # 21 The plant is in MODE 5 with the following conditions:

• Core Alternations are in progress

• It becomes necessary to remove a 125 VDC Station Battery from service Which one of the following is the Technical Specifications implication of removing this battery from service?

The affected 125 VDC Power DISTRIBUTION System ...

A. shall be considered inoperable and the appropriate LCO entered.

B. is operable provided its associated battery charger is operable.

C. is operable provided two independent battery chargers are operable.

D. shall be considered inoperable but is not required in this plant condition.

Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - If a battery is disconnected and only a charger is supplying the bus; the affected 125 VDC Power Distribution System shall be considered inoperable. With a required 125 VDC battery or distribution subsystems inoperable during SDC operations, Core Alts, OPDRVs, moving fuel, etc, either immediately declare inoperable any required features that are dependent on 125 vdc, or immediately suspend all such activities.

B. Incorrect - If a battery is disconnected and only a charger is supplying the bus; the affected 125 VDC Power Distribution System shall be considered inoperable.

C. Incorrect - If a battery is disconnected and only a charger is supplying the bus; the affected 125 VDC Power Distribution System shall be considered inoperable.

D. Incorrect - With a required 125 VDC battery or distribution subsystems inoperable during SDC operations, Core Alts, OPDRVs, moving fuel, etc, either immediately declare inoperable any required features that are dependent on 125 VDC, or immediately suspend all such activities.

OI-302, pgs 4 & 5 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: none Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

12-29-10-is this OK for ROs 6-3-11 changed to mode 5 in stem 6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 215005 2.1.30 Importance Rating 4.4 Conduct of Operations: Ability to locate and operate components, including local controls.

Question: RO Question # 22 The plant is in MODE 1 at 90% power with the following conditions:

• All APRMs are currently OPERABLE

• The A and D APRMs are currently bypassed Due to a maintenance activity, the CRS directs the C APRM be bypassed.

What other APRM, if any, shall be bypassed IAW approved procedures?

A. APRM B shall be bypassed using the APRM bypass switch on the LEFT side of 1C05.

B. APRM B shall be bypassed using the APRM bypass switch on the RIGHT side of 1C05.

C. APRM D shall remain bypassed, can be verified using the APRM bypass switch on the LEFT side of 1C05.

D. APRM D shall remain bypassed, can be verified using the APRM bypass switch on the RIGHT side of 1C05.

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - With C bypassed, the companion APRM that should be bypassed is B APRM. The B APRM is bypassed using APRM bypass switch on the right side of 1C05.

B. Correct - With C bypassed, the companion APRM that should be bypassed is B APRM. The B APRM is bypassed using APRM bypass switch on the right side of 1C05.

C. Incorrect - With C bypassed, the companion APRM that should be bypassed is B APRM. The B APRM is bypassed using APRM bypass switch on the right side of 1C05.

D. Incorrect - With C bypassed, the companion APRM that should be bypassed is B APRM. The B APRM is bypassed using APRM bypass switch on the right side of 1C05.

Technical Reference(s): OI-878.4, P&L 12, NOTE on p11 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

6-9 NRC OK enhancement ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 218000 K1.01 Importance Rating 4.0 Knowledge of the physical connections and/or cause- effect relationships between AUTOMATIC DEPRESSURIZATION SYSTEM and the following: RHR/LPCI: Plant-Specific Question: RO Question # 23 A Loss of Coolant Accident occurred with and the following conditions exist:

• Drywell pressure is currently 10 psig, rising slowly

• ADS has initiated and all 4 ADS valves are open

• RHR Pumps A and C are running on minimum flow

• RHR Pumps B and D will not start

• CS A and B will not start Which one of the following conditions would cause the ADS valves to close?

A. Securing either RHR Pump.

B. Raising RPV level to 65 inches C. Securing both the RHR Pumps D. Reducing RPV pressure to 100 psig Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Either RHR Pump running will provide a permissive for ADS valves to remain open.

B. Incorrect - Clearing the Low Level setpoint will NOT close the SRVs because after the system initiates this signal is bypassed.

C. Correct - Securing both RHR Pumps removes the permissive for the SRVs to open causing them to close.

D. Incorrect - The SRVs will remain open until reactor system pressure lowers to approximately 50 psi above Drywell/Torus pressure, the pilot valve will reseat and the main valve spring pressure will reseat the main disc. In this case approximately 60 psig.

Technical Reference(s): SD-183-1, pg 14 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # DAEC #19343 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 212000 A3.05 Importance Rating 3.9 Ability to monitor automatic operations of the REACTOR PROTECTION SYSTEM including:

SCRAM instrument volume level Question: RO Question # 24 The plant is in MODE 5 with Core Alterations currently in progress. Mode Switch is in REFUEL.

Which one of the following would result in a FULL reactor scram?

A. CRD Scram Discharge Volume high level trip of 60 gallons B. Inadvertent closure of all of the OUTBOARD MSIVs C. Intermediate Range Monitor "A" upscale spike to 120/125 on Range 1 due to undervessel work.

D. Tripping of the Main Turbine at 1C07 using the Turbine Trip pushbutton Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - CRD Scram Discharge Volume High Water Level is sensed in the instrument volume. A level of 60 gallons will result in a full reactor scram.

B. Incorrect - With the plant shutdown for refueling the MSIV isolation scram is bypassed.

C. Incorrect - A single IRM trip would only cause a half scram.

D. Incorrect - With the plant shutdown for refueling the turbine stop valve scram is bypassed.

Technical Reference(s): SD-358, pg 13 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

6-3-11-changed distractor D 6-9 went back to original D. NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 209001 A4.02 Importance Rating 3.5 Ability to manually operate and/or monitor in the control room: Suction valves Question: RO Question # 25 The plant is in MODE 5 with RPV level at the RPV flange in preparation for flood up. Core Spray keylock switch E21A-S16A SUCTION PATH INTERLOCK HS-2103A is placed in the BYPASS position.

What is the bases for placing the switch in the BYPASS position?

This switch A. overrides the automatic opening of the Core Spray suction valves on a system initiation.

B. permits closing the Core Spray suction valve when the CST suction valve is opened.

C. overrides the automatic opening of the Core Spray minimum flow valve when a CST suction valve is open.

D. permits the pump to be run with suction from the CST, with the torus suction path isolated.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The switch has no function related to an automatic initiation.

B. Incorrect - The valves can be repositioned prior to placing the switch in bypass.

C. Incorrect - The switch has no function related to the minimum flow valve.

D. Correct - In order to provide for use of the condensate storage tanks as an alternate suction source, keylocked Core Spray Pump A [B] Suction Path Intlk switches on panel 1C43 [1C44] bypass the loss of suction path interlock when placed in BYPASS. This permits the pumps to be run with suction from the condensate storage tanks, with the torus suction path isolated.

OI-151, Sect. 10, pg 31 Technical Reference(s): (Attach if not previously provided)

SD-151, pgs 9 & 10 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3-11-changed distractor D wording to CST 6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 203000 K2.03 Importance Rating 2.7 Knowledge of electrical power supplies to the following: Initiation logic Question: RO Question # 26 The plant is operating at 100% power when a loss of 120 VAC Instrument Bus 1Y21 occurs.

Which of the following describes the effect of this power loss on the RHR pumps?

A. On the power loss, ONLY RHR Pumps B and D automatically start and operate on minimum flow B. On the power loss, all RHR Pumps automatically start C. If a LPCI initiation signal is received, ONLY A and C RHR pumps would AUTO start D. If a LPCI initiation signal is received, all RHR pumps would AUTO start as designed Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - No pump starts occur.

B. Incorrect - No pump starts occur.

C. Incorrect - RHR logics are cross-divisionalized such that a loss of one 120 VAC Instrument supply does not impact LPCI pump starts.

D. Correct - RHR logics are cross-divisionalized such that a loss of one 120 VAC Instrument supply does not impact LPCI pump starts.

Technical Reference(s): SD-317-1 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 2.1.1.66 2.1.1.7a Learning Objective: 2.1.1.7b (As available) 2.2.1.2 2.3.1.4 Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3-11-believe this is comprehensive because knowledge integration required as shown in the explanation.

6-9 added LO. Leave as LOK Analysis ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 201002 K1.04 Importance Rating 3.5 Knowledge of the physical connections and/or cause- effect relationships between REACTOR MANUAL CONTROL SYSTEM and the following: Rod block monitor: Plant-Specific Question: RO Question # 27 The plant is operating in MODE 1 at 100% power with the following conditions:

• Repairs on "A" Rod Block Monitor have just been completed

• RBM A is removed from BYPASS to accomplish Post Maintenance Testing

• The ROD OUT PERMISSIVE light extinguished and then illuminated again within two seconds

• Annunciator 1C05B (A-6), ROD OUT BLOCK did NOT alarm Which one of the following statements describes the system response to the above?

This condition is ...

A. NOT normal because the A RBM should not null until a new control rod is selected B. normal because "A" RBM generated a rod out inhibit during the null sequence.

C. NOT normal only because the annunciator should have alarmed when the ROD OUT PERMISSIVE light was extinguished.

D. normal because the rod out blocks are bypassed for two seconds to allow the reference APRM gain adjustment during the null sequence.

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect -.Taking the RBM out of BYPASS will initiates a null sequence.

B. Correct - Taking a RBM out of BYPASS initiates a null sequence. RBM trip functions are bypassed during the nulling sequence so no alarm is generated.

C. Incorrect - The RBM trip functions are bypassed during the nulling sequence so no alarm is generated.

D. Incorrect - There is no rod block bypass, the RBM trip functions are bypassed during the nulling sequence so no alarm is generated.

Technical Reference(s): SD-878-5, pg 16 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # LOT Bank 19363 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

6-3 changed distractor A but believe original was OK 6-9 NRC OK with change ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 256000 K2.01 Importance Rating 2.7 Knowledge of electrical power supplies to the following: System pumps Question: RO Question # 28 With the plant operating at full power, the following alarms are received:

• 1C08B A-9, BUS 1A2 LOCKOUT TRIP OR LOSS OF VOLTAGE

• 1C06A C-12, A RX FEED PUMP 1P-1A LOW SUCTION PRESS

• 1C06A C-13, B RX FEED PUMP 1P-1B LOW SUCTION PRESS Which one of the following describes the status of operating Condensate and Feedwater Pumps?

A. ONLY the A Condensate Pump is operating.

B. ONLY the B Condensate Pump is operating.

C. The A Condensate Pump AND the A Feed Water Pump are operating.

D. The B Condensate Pump AND the B Feed Water Pump are operating.

Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Identifies potential misconception of 1P-1A Low Suction Pressure TRIP.

B. Incorrect - Would be true for Bus 1A1 Lockout with potential misconception of 1P-1A Low Suction Pressure TRIP.

C. Correct - Bus 1A2 Lockout de-energizes BOTH Condensate Pump 1P-8B AND Feed Water Pump 1P-1B.

D. Incorrect - Would be true for Bus 1A1 Lockout.

Technical Reference(s): SD-639 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 2007 NRC exam Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

6-3-11-deleted low pressure alarm 6-9-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 290003 K3.01 Importance Rating 3.5 Knowledge of the effect that a loss or malfunction of the CONTROL ROOM HVAC will have on following: Control room habitability Question: RO Question # 29 The plant is operating in MODE 1 at 100% power with the following conditions:

• All LCOs are met Which one of the following is a consequence of prolonged operation the Control Building Ventilation System in the PURGE mode?

The PURGE mode ...

A. bypasses the heating and cooling coils resulting in loss of Control Building temperature control.

B. isolates the outside air intake lowering Control Building pressure below atmospheric pressure.

C. ventilation flow bypasses the Cable Spreading and Battery Rooms which may result in having to declare the Batteries inoperable.

D. closes the Control Room Recirculation Damper which could result in more rapid buildup of radiological or toxic chemical concentrations.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - When HS 6107 is placed in the Fresh Air mode of operation, the Control Room Recirculation Damper DO6109 fully closes, this mode does not bypass the heating and cooling and temperature is not a concern.

B. Incorrect - Damper Operator DO6106A(B) maintains mixing plenum (supply fan suction)

.25"wg greater than outside pressure.

C. Incorrect - Placing the Control Building Ventilation system in the PURGE mode does not bypass the Cable Spreading and Battery Rooms.

D. Correct - When HS 6107 is placed in the Fresh Air mode of operation, the Control Room Recirculation Damper DO6109 fully closes. The basis for use of the fresh/auto (purge) mode is at the discretion of the OSM/CRS for comfort in the control room only. If the control building ventilation is operated in purge mode for extended periods, and a radiological or toxic chemical event were to occur, the higher intake flow rate in PURGE mode could result in more rapid buildup of radiological or toxic chemical concentrations than has been assumed in the safety analysis.

OI-730, pg 6 Technical Reference(s): SD-730- pg 37 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 233000 K4.06 Importance Rating 2.9 Knowledge of FUEL POOL COOLING AND CLEAN-UP design feature(s) and/or interlocks which provide for the following: Maintenance of adequate pool level Question: RO Question # 30 Which one of the following is:

(1) The Minimum Technical Specifications required Fuel Pool water level? AND (2) How is this level controlled?

A. (1) 36 ft.

(2) A series of weirs controls the Fuel Pool minimum level and the maximum level is controlled by manually throttling makeup water.

B. (1) 23 ft. above the top of the fuel racks.

(2) A series of weirs maintains a specific level and the maximum level is controlled by automatic level control of the Fuel Pool Skimmer Surge Tank.

C. (1) 36 ft.

(2) A series of weirs maintains a specific level and the maximum level is controlled by automatic level control of the Fuel Pool Skimmer Surge Tank.

D. (1) 23 ft. above the top of the fuel racks.

(2) A series of weirs controls the Fuel Pool minimum level and the maximum level is controlled by manually throttling makeup water.

Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - The Tech Spec limit for FP level is >36 ft. A series of weirs controls the Fuel Pool minimum level the maximum level is controlled by manually throttling makeup water IAW OI-435, Sect 6.0.

B. Incorrect - This 23' above the top of fuel is the Technical Specifications for Reactor Pressure Vessel (RPV) Water Level during Refueling Operations above the fuel in the RPV. There is no automatic level control of the Fuel Pool Skimmer Surge Tank C. Incorrect - There is no automatic level control of the Fuel Pool Skimmer Surge Tank D. Incorrect - This 23' above the top of fuel is the Technical Specifications for Reactor Pressure Vessel (RPV) Water Level during Refueling Operations above the fuel in the RPV.

1C04B, A-4 Technical Reference(s): (Attach if not previously provided)

OI-435, Sect 6.0.

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-3-11 -changed to on all distractors 6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 201006 K5.01 Importance Rating 3.3 Knowledge of ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following: Minimize clad damage if a control rod drop accident (CRDA) occurs: P-Spec(Not-BWR6)

Question: RO Question # 31 Which one of the following describes the design basis function of the Rod Worth Minimizer?

It enforces A. rod withdrawal with a programmed control rod sequence to limit the power excursion to prevent rapid dispersal of the fuel in the event of a Control Rod Drop Accident (CRDA)

B. control rod sequences designed to prevent exceeding the Minimum Critical Power Ratio when Reactor power is below 21.7% Rated Thermal Power C. programmed rod movement that minimizes individual control rod worth to prevent exceeding the Maximum Extended Load Limit Analysis (MELLA) while in MODE 2 D. control rod sequences to limit the rate of heat production to < 280 calories/gram of fuel during control rod withdrawal when reactor power is > 21.7%.

Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - Since the worth of an individual rod is highly dependent on core power distribution, rod sequence control provides a means of restricting the maximum reactivity insertion that could occur in a CRDA. The principal function of the NUMAC RWM is to limit rod motion such that high worth rods are not created, thereby limiting the maximum reactivity which could be added due to a control rod drop accident.

B. Incorrect - This is not a design function, the RWM does ensure that fuel operating limits are not exceeded and that the possibility of a high notch worth scram occurring is minimized.

C. Incorrect - This is not a design function, the RWM does ensure that fuel operating limits are not exceeded and that the possibility of a high notch worth scram occurring is minimized.

D. Incorrect - The RWM limits the rate of heat production to < 280 calories/gram of fuel during rod DROP accident NOT a control rod withdrawal. And the power level is when reactor power is <10%.

Technical Reference(s): SD-878.8, pg 4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5,6 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Design, components, and functions of reactivity control mechanisms and instrumentation.

Comments:

6-3-11-OK for ROs (is there a LO?), added 55.41 (5) 6-9-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 245000 K6.10 Importance Rating 2.8 Knowledge of the effect that a loss or malfunction of the following will have on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS: Lube oil system Question: RO Question # 32 The plant is operating in MODE 1 at 100% power with the following conditions:

• Turbine Building NSPEO reports that a very large lube oil leak has developed near the Main Generator

• Subsequent to the report alarm 1C07A A-7, TURBINE LUBE OIL BEARING HEADER LO PRESSURE activates

• The Turbine Building NSPEO reports that he cannot maintain Lube Oil Tank level Which actions are required by AOP 693, Main Turbine/EHC Failures?

The ___(1)___ and the condenser vacuum shall be ___(2)___.

A. (1) Reactor will be scrammed then Main Turbine manually tripped (2) broken B. (1) Main Turbine will be tripped, and automatic Reactor scram verified (2) broken C. (1) Reactor will be scrammed then Main Turbine manually tripped (2) maintained D. (1) Main Turbine will be tripped, and automatic Reactor scram verified (2) maintained Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - the reactor is scrammed, then the turbine is tripped, MSIVs are closed to facilitate breaking Main Condenser vacuum B. Incorrect - the turbine is tripped before the reactor is scrammed, MSIVs are closed to facilitate breaking Main Condenser vacuum C. Incorrect - the reactor is scrammed, then the turbine is tripped, MSIVs are closed to facilitate breaking Main Condenser vacuum D. Incorrect - the turbine is tripped before the reactor is scrammed, MSIVs are closed to facilitate breaking Main Condenser vacuum Technical Reference(s): AOP-693 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # # 20729 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-9 NRC OK with Change - still unsat, but fixed 6-14 realized that A/C B/D pairs were not different, fixed part (2) answers to have questions different.

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 202001 A1.02 Importance Rating 3.4 Ability to predict and/or monitor changes in parameters associated with operating the RECIRCULATION SYSTEM controls including: Jet pump flow Question: RO Question # 33 The plant is conducting a startup with the following conditions:

• The reactor is critical

• Reactor power is approximately 1%, 50 on range 8 of IRMs

• Reactor pressure is 950 psig

• The A Recirculation Pump has just tripped With these plant conditions; (1) Which one of the following indications must the Reactor Operator monitor?

(2) What is indicated by these indications?

A. (1) Excessive noise on the jet pump dP indicators (2) Jet pump cavitations B. (1) High flow indication on the operating loops jet pumps (2) Jet pump cavitations C. (1) Excessive noise on the jet pump dP indicators (2) Cavitation of the operating recirculation pump D. (1) High flow indication on the operating loops jet pumps (2) Cavitation of the operating recirculation pump Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - IAW with OI-264, P & L 5and 10, at rated temperature and low reactor power (less than 2%), avoid single loop operation, even at minimum speed. If single loop operation is necessary for short periods of time, monitor jet pump flow to ensure cavitation does not occur. Jet pump cavitation is indicated by excessive noise on the jet pump dP indicators. In this question the plant is below 2% power (Range 8 0 on the IRMs and at rated pressure.

B. Incorrect - Jet pump cavitation is indicated by excessive noise on the jet pump dP indicators.

C. Incorrect - Recirc Pump cavitation is indicated by excessive vibration and sudden drop in pump discharge pressure and flow D. Incorrect - Recirc Pump cavitation is indicated by excessive vibration and sudden drop in pump discharge pressure and flow OI-264, P & Ls 5 and 10, pgs 4 &

Technical Reference(s): (Attach if not previously provided) 5 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

5-09-11, Revised question 6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 272000 A2.15 Importance Rating 2.5 Ability to predict the impacts of the following on the RADIATION MONITORING SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Maintenance operations Question: RO Question # 34 The plant is operating in MODE 1 at 100% power with the following conditions:

• The FUEL POOL EXHAUST RADIATION MONITOR RIS-4131A Mode Switch is taken out of the OPERATE position by an I&C Technician (1) Which one of the following initiations will occur?

(2) What action is required?

A. (1) Only the "A" Standby Gas Treatment system will initiate (2) IAW OI-170, SBGT, verify the proper operation of SBGT B. (1) Only the "A" Standby Gas Treatment system will initiate.

(2) IAW IPOI-7, Special Operations, verify the automatic isolation of the Secondary Containment ONLY.

C. (1) Both Standby Gas Treatment systems will initiate.

(2) IAW IPOI-7, Special Operations, verify the automatic isolation of the Secondary Containment ONLY D. (1) Both Standby Gas Treatment systems will initiate.

(2) IAW OI-170, SBGT, verify the proper operation of SBGT, then it is required to shutdown one train of SBGT Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct -. The Pool exhaust high radiation of 8 mr/hr or mode switch out of operate will initiate the A SBGT train ONLY. Primary and Secondary Containment will automatically isolate. Since the SBGT System started on error, the system operation is verified, then the SBGT system can be returned to STANDBY.

B. Incorrect -. Primary AND Secondary Containment will automatically isolate, and will be verified via IPOI 7.

C. Incorrect - ONLY the A SBGT train will automatically start, and Primary AND Secondary Containment will automatically isolate D. Incorrect - ONLY the A SBGT train will automatically start. It is NOT required to shutdown one train of SBGT.

OI-170, pgs 8 and 9 Technical Reference(s): SD-170 (Attach if not previously provided)

SD 959.1, page 21 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3-11-changed distractors as above. Changed b explanation to incorrect.

6-9-11-NRC OK. Need to fix explanations. Changed question to balance per NRC request.

This made two correct answers, so changed question distractors to make only one correct answer.

6/14/1 - deleted part of A(2), which was wrong (two sbgt actions, vs one running)

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 286000 A3.04 Importance Rating 3.2 Ability to monitor automatic operations of the FIRE PROTECTION SYSTEM including: System initiation Question: RO Question # 35 The plant is operating in MODE 1 at 100% power with the following conditions:

• A and B Cooling Towers are in service

• A small nitrogen leak inside the shroud of the E Cooling Tower cell causes the deluge for the E and F Cells to initiate Which one of the following describes the effect of this initiation on Cooling Tower Fan operation?

The cooling tower fans will automatically ...

A. trip if running in FWD, but remain running if running in REVERSE B. remain running unless high temperatures are confirmed by local temperature switches C. trip if running in FWD or REVERSE. Taking the handswitch on 1C06 to STOP will reset the logic and allow the fan to be reset with no other operator actions D. trip if running in FWD or REVERSE. The cooling tower deluge must be isolated and then reset in order to restart the fans Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Activation of the Cooling Tower Deluge System automatically shuts off the associated tower fans.

B. Incorrect - Activation of the Cooling Tower Deluge System automatically shuts off the associated tower fans.

C. Incorrect - Activation of the Cooling Tower Deluge System automatically shuts off the associated tower fans when a pressure switch reads 6 psig pressure in the deluge system. The fan will not start until the pressure switch resets, meaning no pressure.

The procedure isolates the deluge, then drains the deluge piping.

D. Correct - Activation of the Cooling Tower Deluge System automatically shuts off the associated tower fans when a pressure switch reads 6 psig pressure in the deluge system. The fan will not start until the pressure switch resets, meaning no pressure.

The procedure isolates the deluge, then drains the deluge piping.

OI-513, pg 4 Technical Reference(s): (Attach if not previously provided)

ARP 1C06A A-5 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-9 Added word fan to stem 6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 215002 A4.01 Importance Rating 2.8 Ability to manually operate and/or monitor in the control room: IRM/RBM recorder/switch:

BWR-3,4,5 Question: RO Question # 36 The plant is in Mode 2.

With the B IRM bypassed, which set of the following B IRM indications remains available?

1 - B IRM 1C05 indicating lamps on the Reactor Control Benchboard (EXCEPT bypass light) 2 - IRM "B" inputs to the IRM recorder 3 - "B" IRM outputs to the annunciators 4 - "B" IRM channel inputs to SPDS 5 - 1C36 meter indications for the "B" IRM A. 1, 3, 4 B. 2, 4, 5 C. 1, 2, 4 D. 2, 3, 5 Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The IRM outputs to the indicating lamps on the Reactor Control Benchboard and IRM outputs to the annunciator are defeated.

B. Correct - When an IRM channel is bypassed, the following IRM functions are defeated:

a. The IRM UPSCALE trip to Reactor Protection System.

b. The IRM associated trips to the rod withdrawal block circuits of the Reactor Manual Control System.

c. The IRM outputs to the annunciator and sequence recorder.

d. The IRM outputs to the indicating lamps on the Reactor Control Benchboard. The Retract Permit Lamp will remain ON as long as the IRM channel is bypassed and the IRM detector is not full out.

C. Incorrect - The IRM outputs to the indicating lamps on the Reactor Control Benchboard are defeated.

D. Incorrect - The IRM outputs to the annunciator are defeated.

OI-878.2, NOTE pg 12 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # # 20455 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and function of reactivity control mechanisms and instrumentation.

Comments:

6-3-11-revised stem 6-9 NRC OK, removed 1 choice (previously #5)

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 268000 2.4.21 Importance Rating 4.0 Emergency Procedures / Plan: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Question: RO Question # 37 The plant is operating in MODE 1 at 100% power with the following conditions:

• Annunciator A-2 REACTOR BLDG SOUTH EAST AREA FLOOR DRAIN LEVEL HIGH alarms at panel 1C147, RB Floor Drain System Control

• Annunciator B-4 AREA WATER LEVELS ABOVE MAX NORMAL alarms at panel 1C14A, EOP Annunciators

• An operator reports from 1C21 that SE Corner Room level is slightly greater than 2 inches and rising very slowly.

• SANSOE reports from the SECR mezzanine that there is water on the floor and he will try to locate the leak Which one of the following procedures:

(1) Shall be reported to the CRS as a possible entry, and (2) What are the required actions A. (1) EOP 1, RPV CONTROL (2) Scram the reactor and control level, pressure, reactor power.

B. (1) EOP 3, SECONDARY CONTAINMENT CONTROL (2) Contact the Plant Chemist and have him sample the water prior to draining it to the Reactor Building Floor Drain Sump.

C. (1) EOP 1, RPV CONTROL (2) Contact the Radwaste Operator and have him pump down the Reactor Building Floor Drain Sump.

D. (1) EOP 3, SECONDARY CONTAINMENT CONTROL (2) Have the Radwaste Operator open the affected valve to drain the area, and operate sump pumps as necessary.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The greater than max normal water level is an entry into EOP 3, not EOP 1.

B. Incorrect - There is no requirement to sample the water and time should not be spent in the EOP sampling the discharge of water from this area is required.

C. Incorrect - The greater than max normal water level is an entry into EOP 3, not EOP 1.

D. Correct - SE Corner Room level is slightly greater than 2 inches is above the Max Normal Operating Limit for the SE corner Room which requires an entry into EOP-3.

The EOP requires operating available sump pumps to restore and maintain water level below the Max Normal Operating Limit Technical Reference(s): EOP-3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-3 changed part 1 of A and C to EOP 1, corrected name of EOP 3 6-9-11-NRC OK with Change ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 216000 K5.10 Importance Rating 3.1 Knowledge of the operational implications of the following concepts as they apply to NUCLEAR BOILER INSTRUMENTATION: Indicated level versus actual vessel level during vessel heatups or cooldowns Question: RO Question # 38 The plant is shutting down following a steam leak in the Drywell, the following conditions exist:

• Drywell temperature has raised to 350°F

• RPV pressure is stable at 100 psig

• The Action Is Required area of EOP Graph 1 has been entered Which of the following statements is correct regarding the RPV level instruments?

A. RPV actual level and indicated level will be equal under these conditions.

B. RPV actual level may be higher than indicated level due to boiling in the RPV.

C. RPV indicated level may be higher than actual level due to reference leg heating.

D. RPV level may only be read on the Floodup and Wide Range Yarway instruments.

Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Because of the lower density of the reference leg indicated level will be higher than actual level.

B. Incorrect - With the reactor stable the temperature (~338°F) is above drywell temperature so no boiling in the reference or variable legs will occur. Indicated level may be higher than actual level due to reference leg heating.

C. Correct - With the reactor stable, an increase in containment temperature would cause an increase in the temperature of the reference leg. This would lower the density making the reference leg "lighter". Because level is derived by a dp cell measuring the difference in weight, the decrease in the weight of the reference leg would cause a loss of inventory from the reference legs which would result in erroneously high indications D. Incorrect - the Floodup and Wide Range Yarway instruments will be affected although they may still be used for level indication the GMAC level indicator also provide level indication.

DAEC EOP 2 Bases Document, EOP Curves and Limits, pgs. 81-Technical Reference(s): (Attach if not previously provided) 83, SD-880, pgs. 30-32,44-45 Proposed References to be provided to applicants during examination: None RO 95.00.00.14 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 2 55.43 General Design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

6-9-11-NRC OK with new question. Needed to add additional bullet and changed DW/T to prevent possible two correct answers.

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295023 AK1.01 Importance Rating 3.6 Knowledge of the operational implications of the following concepts as they apply to REFUELING ACCIDENTS: Radiation exposure hazards Question: RO Question # 39 The plant is in MODE 5, REFUELING, and Core Alterations in progress.

• RPV level begins to lower unexpectedly In accordance with Technical Specifications which of the following is the MINIMUM acceptable water level above the top of the irradiated fuel assemblies seated within the RPV?

A. 201 B. 23 C. 36 D. 37.5 Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Below the actual required limit by TS B. Correct -IAW TS 3.9.6 C.

Incorrect - This is the TS for normal fuel pool level D. Incorrect - This is the normal Fuel Pool Level Technical Reference(s): TS 3.9.6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-3 new question based on AOP 981 6-9 revised ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295025 EK1.06 Importance Rating 3.5 Knowledge of the operational implications of the following concepts as they apply to HIGH REACTOR PRESSURE : Pressure effects on reactor water level Question: RO Question # 40 The plant was operating in MODE 1 at 98% power due to coastdown with the following conditions:

• A Loss of Vacuum event occurred

• A manual Reactor Scram was inserted

• All Control Rods are Full In

• Bypass Valves have failed closed.

• RPV Water Level is being maintained by Feedwater.

• Low Low Set is NOT working Under these conditions stabilizing reactor pressure less than 1055 psig will ___(1)___ and

___(2)___.

A. (1) avoid repeated operation of the SRVs on high reactor pressure (2) prevent SRV damage due to two phase flow B. (1) allow the operator to manually reset the ATWS ARI/RPT logic if it initiated on high reactor pressure (2) prevent potential SRV damage due to the frequent cycling C. (1) avoid repeated operation of the SRVs on high reactor pressure (2) assist in maintaining RPV level below the high level trip setpoint D. (1) allow the operator to manually reset the ATWS ARI/RPT logic if it initiated on high reactor pressure (2) assist in maintaining RPV level below the high level trip setpoint Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - This will avoid repeated operation of the SRVs on high reactor pressure however the concern with SRV openings is RPV level swell.

B. Incorrect - manual reset of the scram would be possible NOT ATWS ARI/RPT logic.

The concern with SRV openings is RPV level swell.

C. Correct - Per EOP 1 Bases - Swell resulting from SRV actuation may result in high level trips of steam driven systems even if level is maintained low in the normal band. It may then be necessary to define a wider control band to maintain level below the high level trip setpoint. Bases for RC/P-4 step Stabilize RPV pressure Below 1055 psig - The direction to stabilize RPV pressure in Step RC/P-4 means to limit changes in RPV pressure (both increases and decreases) to within as small a band as possible.

Controlling RPV pressure below this value avoids SRVs lifting on high pressure and allows the scram logic to be reset (provided no other scram signal exists).

D. Incorrect - manual reset of the scram would be possible NOT ATWS ARI/RPT logic.

EOP 1 bases page 24 and 55 Technical Reference(s): (Attach if not previously provided)

(Rev 14)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6-3-11-changed all part (2)s and stem.

6-9 NRC OK with change ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295024 EK1.01 Importance Rating 4.1 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE : Drywell integrity: Plant-Specific Question: RO Question # 41 The plant was operating at rated power when a DBA LOCA occurred.

Under these conditions, ___(1)___ could cause the drywell to exceed its ___(2)___ design pressure limit.

A. (1) a Torus to Drywell Vacuum Breaker failing OPEN (2) internal.

B. (1) a Torus to Drywell Vacuum Breaker failing CLOSED (2) external C. (1) a Reactor Building to Torus Vacuum Breaker failing OPEN (2) external D. (1) a Reactor Building to Torus Vacuum Breaker failing CLOSED (2) internal Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - IAW SD 959 - Containment Characteristics after LOCA with Torus /Drywell Vacuum Breaker Failed Open - Steam flows from the drywell to the torus through the vacuum breaker equalizing the pressure. The steam is not forced through the downcomers and up through the water, but instead is dumped on the surface of the water in the torus. As a result, the drywell pressure will probably exceed design pressure.

B. Incorrect - In this condition, drywell pressure could lower and cause the Torus to Drywell differential pressure to exceed 2 psid.

C. Incorrect - correct if the vacuum breaker failed closed D. Incorrect - IAW SD 959 page 25, if a reactor building to torus vacuum breaker were to be failed closed in the case of a DBA, there would be little effect. The purpose of the reactor building to torus vacuum breakers is to ensure that neither the torus nor drywell exceed their external pressure limit.

Technical Reference(s): SD 959 rev 4 page 24 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295005 AK2.05 Importance Rating 2.6 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: Extraction steam system Question: RO Question # 42 The plant is operating in MODE 1 at 100% power with the following conditions:

• A Main Turbine trip occurs How is the extraction steam system affected?

The High Pressure Extraction Drain to Condenser, CV-1237 will fail ___(1)___ due to the Extraction Relay Dump Valve ___(2)___.

A. (1) open (2) opening B. (1) closed (2) closing C. (1) open (2) closing D. (1) closed (2) opening Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - CV-1237 fails OPEN due to Extraction Relay Dump Valve opening on loss of EHC pressure due to the turbine trip.

B. Incorrect - The High Pressure Extraction Drain to Condenser, CV-1237, opens, as does the Extraction Relay Dump Valve.

C. Incorrect -On any Main Turbine trip, High Pressure Extraction Drain to Condenser CV-1237 opens due to Extraction Relay Dump Valve opening.

D. Incorrect - The High Pressure Extraction Drain to Condenser, CV-1237, opens and the Relay Dump Valve opens.

SD 646 Rev.10 page 33 Technical Reference(s): (Attach if not previously provided)

SD 693.2 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3 changes valve name in stem 6-9 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295003 AK2.04 Importance Rating 3.4 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF A.C. POWER and the following: A.C. electrical loads Question: RO Question # 43 The plant is operating in MODE 1 at 35% power with the following conditions:

• The "A" Circ Water Pump is in operation

• The A Cooling Tower is in operation Assuming no operator action, which of the following conditions would result in a trip of the A Circ Water Pump or indicate the pump tripped?

A. Circ Water Pit level lowering to 13 ft B. Losing 1Y11, Instrument AC Division 1 C. Losing 1Y23, 120 VAC Uninterruptible power supply D. 1C06A, CIRC WATER PUMP 1P-4A HI VIBRATION (B-10) alarms Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

Incorrect - There is an administrative limit of 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> of operation with Circ Pit level A: below 13 ft Correct - Loss of 1Y11 power will cause KY-4201 to be deenergized allowing stored energy in the accumulators to be released and close HO-4201 which trips 1P-4A.

B: None of the malfunctions listed is a direct trip of a Circ Pump. All require knowledge of system interactions Incorrect - Losing 1Y23 has no effect on Circ Water pumps, but if the candidate C: confuses the 1Y11 action, this is a plausible choice.

Incorrect - There are no automatic actions associated with the Circ Water Pump High D: Vibration alarm.

OI-442 "Circulating Water System" Technical Reference(s): Rev. 81, P&L #7 (Attach if not previously provided) 1C06A B-10 Proposed References to be provided to applicants during examination: None Learning Objective: 32.02.02.02 (As available)

Question Source: Bank # WTS 10375 Modified Bank # (Note changes or attach parent)

New Question History:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-10 NRC OK with change enhancement not unsat ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 295001 AK2.03 Importance Rating 3.6 Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION and the following: Reactor water level Question: RO Question # 44 The plant is operating in MODE 1 at 100% power with the following conditions:

• The B Reactor Recirculation Pump tripped

• All systems responded as designed Which of the following describes the INITIAL reactor water level response and why?

Indicated reactor water level will ___(1)___ due to the ___(2)___.

A. (1) RISE (2) collapse of steam voids B. (1) LOWER (2) lack of coolant velocity to sweep voids into the steam separator C. (1) RISE (2) displacement of water by increased steam voiding D. (1) LOWER (2) initial delay in feedwater control system response Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - steam voiding would increase B. Incorrect - steam voiding would increase C. Correct - the trip of the pump would result in more steam voiding. RPV would increase until the FW control system restored level to the normal value D. Incorrect - level would increase due to increased voiding GFES Chapter 8, Operational Physics, discussion on RR flow Technical Reference(s): and Reactor Power (discussion is (Attach if not previously provided) to increase RR flow, this question is reversed)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTS 1109 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 600000 AK3.04 Importance Rating 2.8 Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE:

Actions contained in the abnormal procedure for plant fire on site Question: RO Question # 45 The plant is operating in MODE 1 at 100% power with the following conditions:

• A RHR loop is tagged out of service for maintenance

• A fire has been verified in the turbine building, in Fire Area TB1 Which of the following is an action that is required IAW AOP 913, Fire, and why?

Dispatch an NSPEO to ____.

A. manually close MO-1905, RHR LOOP B LPCI INBD INJECT ISOL if it spuriously opens to prevent RPV injection when not required.

B. manually open MO-1905, RHR LOOP B LPCI INBD INJECT ISOL if only B RHR is available to ensure an RPV injection path.

C. manually open V-19-48, RHR LOOP CROSSTIE to ensure an RPV injection supply if only B RHR is available for RPV injection.

D. manually open BOTH V-19-48, RHR LOOP CROSSTIE and MO-1905, RHR LOOP B LPCI INBD INJECT ISOL to ensure an RPV injection supply if only B RHR is available for RPV injection.

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - no actions listed in AOP 913 to manually close the valve.

B. Correct - IAW AOP 913 Path TB1 continuous recheck statement C. Incorrect - the direction is to CLOSE the V-19-48 valve (RB3 Continuous Recheck Statement, page 83)

D. Incorrect - the direction is to CLOSE the V-19-48 valve (RB3 Continuous Recheck Statement, page 83)

AOP 913 Path TB1 continuous Technical Reference(s): (Attach if not previously provided) recheck statement Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295037 EK3.05 Importance Rating 3.2 Knowledge of the reasons for the following responses as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN: Cold shutdown boron weight: Plant-Specific Question: RO Question # 46 Which of following describes why achieving COLD SHUTDOWN BORON WEIGHT is desired during EOP-ATWS mitigation actions?

To assure the reactor will remain shutdown _____.

A. prior to raising RPV level to 170 to 211.

B. irrespective of control rod position and with RPV water level at a minimum of -25.

C. to allow a reactor cooldown to begin.

D. with RPV water level at a minimum of -25.

Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - this is the concept of Hot Shutdown Boron Weight B. Incorrect - this partially defines Hot Shutdown Boron Weight. RPV level must be in the normal band C. Correct - IAW EOP ATWS Bases, page 68 - Injection of the Cold Shutdown Boron Weight (CSBW) of boron into the RPV ensures that the reactor is shutdown and will remain shutdown. The CSBW is the least weight of soluble boron which, if injected into the RPV and mixed uniformly, will maintain the reactor shutdown under all conditions.

D. Incorrect - this partially defines Cold Shutdown Boron Weight but with the incorrect RPV level.

EOP ATWS Bases Rev.14 page Technical Reference(s): (Attach if not previously provided) 68 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6-10-11-NRC OK with change. Not unsat, enhanced ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295019 AK3.02 Importance Rating 3.5 Knowledge of the reasons for the following responses as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Standby air compressor operation Question: RO Question # 47 The plant is operating in MODE 1 at 100% power with the following conditions:

• 1K1 is in STANDBY mode

• A loss of Instrument Air header pressure occurs

• Instrument Air header pressure is 90 psig and lowering slowly Which one of the following is:

(1) The reason the Backup Air Compressor 1K1 starts at this time?

(2) What system will supply Backup Air Compressor 1K1 cooling?

A. (1) To supply ONLY the Instrument Air Header pressure.

(2) Compressor Cooling Water System B. (1) To supply BOTH the Instrument & Service Air Headers (2) Compressor Cooling Water System C. (1) To supply ONLY the Instrument Air Header pressure.

(2) Well Water System D. (1) To supply BOTH the Instrument & Service Air Headers (2) Well Water System Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - initially both service and instrument air headers are supplied.

B. Incorrect - The 1K1 is supplied by Well water.

C. Incorrect - initially both service and instrument air headers are supplied.

D. Correct - Unless header pressure drops to 82 psig, both headers are supplied. The well water system is the primary cooling water medium for the 1K1 AOP 518 Technical Reference(s): (Attach if not previously provided)

SD 518 Rev 8. pages 13,14,24,27 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295018 AA1.02 Importance Rating 3.3 Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: System loads Question: RO Question # 48 The plant is operating at rated power. The A SBDG is in service for a scheduled surveillance test.

Then, a loss of all River Water Supply (RWS) Pumps occurs.

The plant is manually scrammed and the initial actions of IPOI-5 are completed successfully.

Which of following describes RWS system loads that are DIRECTLY impacted and an action required IAW AOP 410, Loss of River Water Supply.

Monitor ___(1)___ system loads and ___(2)___.

A. (1) ESW, RHRSW and GSW (2) Secure the running SBDG B. (1) Circ Water, RHRSW, and Fuel Pool Cooling (2) Secure the running SBDG C. (1) ESW, RHRSW and GSW (2) Open the Circ Water Inlet to Blowdown Line valve MO-4253 to maintain Circ Water Pit inventory.

D. (1) Circ Water, RHRSW, and Fuel Pool Cooling (2) Open the Circ Water Inlet to Blowdown Line valve MO-4253 to maintain Circ Water Pit inventory.

Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - IAW AOP 410 page 4 step 7 - Shutdown any SBDG not required to ensure one Essential Bus is energized and/or required to ensure adequate core cooling.

IAW SD 410 - RWS Purpose - to provide makeup water from the Cedar River for the Circulating Water System, GSW, RHRSW, ESW, Fire System and Radwaste Dilution Systems to replace that which is lost due to evaporation, blowdown and normal uses.

B. Incorrect - Fuel Pool Cooling is not directly impacted by this loss. It is cooled by RBCCW C. Incorrect - The Circ Water Inlet to Blowdown Line valve MO-4253 is required to be CLOSED.

D. Incorrect - Fuel Pool Cooling is not directly impacted by this loss. It is cooled by RBCCW. The Circ Water Inlet to Blowdown Line valve MO-4253 is required to be CLOSED.

AOP 410 Rev.14 page 4 Technical Reference(s): (Attach if not previously provided)

SD 410 - system purpose Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

6-10 NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 700000 AA1.04 Importance Rating 4.1 Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Reactor controls.

Question: RO Question # 49 The plant is operating in MODE 1 at 100% power with the following conditions:

• ITC Midwest notifies the Main Control Room of a degraded offsite power condition

• 1A3 and 1A4 bus voltage is continuing to degrade toward a trip condition

• 1A3 and 1A4 have not yet tripped Which of the following is required IAW AOP 304 - Grid Instability?

A. (1) Start the SBDGs (2) Parallel and load the Essential Buses (3) Reduce Recirc to 27 mlbm/hr Flow (4) Scram the reactor B. (1) Reduce Recirc to 27 mlbm/hr Flow (2) Scram the reactor (3) Start the SBDGs (4) Parallel and load the Essential Buses before the 1A3 and 1A4 bus supply breakers trip C. (1) Reduce Recirc to 27 mlbm/hr Flow (2) Scram the reactor (3) Do not attempt to start and load the SBDGs (4) Continue to monitor for Grid Instabilities D. (1) Start the SBDGs (2) Do NOT parallel and load the Essential Buses (3) Continue to monitor for Grid Instabilities (4) If the 1A3 and 1A4 trip, verify the SBDGs load their respective buses and the Reactor Scrams Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - would not start the SBDGs with degraded conditions B. Incorrect - would continue to monitor grid instability and continue with IPOI 5 actions.

would not start the SBDGs.

C. Correct - IAW AOP 304 Caution - It is not appropriate to manually start and load a SBDG during degraded grid conditions. Followup action 1.b. - IF It appears that busses 1A3 and 1A4 will trip due to degrading grid conditions. Reduce Recirc to 27 mlbm/hr and Flow Scram the reactor.

D. Incorrect - would not start the SBDGs with degraded conditions Technical Reference(s): AOP 304 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295006 AA1.02 Importance Rating 3.9 Ability to operate and/or monitor the following as they apply to SCRAM: Reactor water level control system Question: RO Question # 50 IPOI-5, Reactor Scram, has been entered and plant conditions are as follows:

• Level setback pushbutton has been depressed

• Scram choreography is complete

• The Feedwater Master Controller, LC-4577, is in AUTO

• RPV level has risen to 175 inches and is stable The CRS directs that RPV level be returned and remain in the green band (186 to 195).

Which one of the following describe actions required to return reactor water level to the normal band IAW IPOI-5, Reactor Scram?

A. Adjust the Feedwater Master Controller LC-4577 in AUTO until reactor level is restored to the green band.

B. Place the Feedwater Master Controller, LC-4577, to MANUAL and adjust flow to return level to the green band. LC-4577 should remain in MANUAL.

C. Reset the Setpoint Setback by depressing the reset pushbutton on 1C05 and then adjusting the Feedwater Master Controller LC-4577 AUTO setpoint until level is in the green band.

D. Place the "A" and "B" Feedwater Regulating Valve Controllers in MANUAL and adjust flow until level is restored to the green band. Then place those controllers back in AUTO.

Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - IAW IPOI 5 - Use any or all of the following techniques as necessary to control RPV level: After RPV level starts to rise as indicated on the wide range Yarways, then place Master Feed Reg controller LC-4577 in MANUAL and close the Feed Reg valves. Restore LC-4577 back to AUTO after RPV level stabilizes.

B. Incorrect - The minimum actions would be to leave the controller in AUTO, and the procedure requires the controller be set back to AUTO.

C. Incorrect - The Feedwater Master Controller, LC-4577, must be in manual to take the setback circuit out of the level control system .

D. Incorrect - not required to place the FRV controllers in manual Technical Reference(s): IPOI 5 Rev 54 step 3.2 (4) a. (Attach if not previously provided)

Proposed references to be provided to applicants during examination: None Learning Objective: RO-45.05.01.05-05 (As available)

Question Source: Bank # 20086 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-10 NRC OK with changes. Enhanced, not unsat ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295030 EA2.02 Importance Rating 3.9 Ability to determine and/or interpret the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: Suppression pool temperature Question: RO Question # 51 In a LOCA event, which of the following is a concern if a Torus Water lowered to a level of 5.8 feet?

(1) What is the specific equipment issue at this elevation?

(2) What are the implications of this equipment being uncovered?

A. (1) The HPCI Turbine Exhaust will become uncovered (2) This will directly pressurize the torus. The consequences of continuing to operate HPCI may result in failure of the primary containment from over pressurization B. (1) The HPCI Turbine Exhaust will become uncovered (2) To ensure that steam discharged from the drywell into the torus following a primary system break will be adequately condensed. If a primary system break were to occur with torus water level below the bottom of the HPCI Turbine Exhaust, pressure suppression capability would be unavailable and torus pressure could exceed the Primary Containment Pressure Limit.

C. (1) The RCIC Turbine Exhaust will become uncovered (2) This will directly pressurize the torus. The consequences of continuing to operate RCIC may result in failure of the primary containment from over pressurization D. (1) The RCIC Turbine Exhaust will become uncovered (2) To ensure that steam discharged from the drywell into the torus following a primary system break will be adequately condensed. If a primary system break were to occur with torus water level below the bottom of the RCIC Turbine Exhaust, pressure suppression capability would be unavailable and torus pressure could exceed the Primary Containment Pressure Limit.

Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - EOP 2 Bases Step TL/6 - (1) A torus level of 5.8 feet corresponds to the HPCI turbine exhaust elevation.

(2) Operation of the HPCI system with its exhaust device not submerged will directly pressurize the torus. HPCI operation is therefore secured when torus level cannot be maintained above 5.8 feet to preclude pressurizing the torus. The consequences of not doing so may result in failure of the primary containment from over pressurization. Thus, HPCI must be secured irrespective of adequate core cooling concerns.

B. Incorrect - (1) The HPCI turbine exhaust level is 5.8 feet (correct), however (2) the discussion is the bases discussion for the 7.1 ft torus level.

C. Incorrect - (1) The RCIC turbine exhaust is at the approximate same level, but the RCIC is not tripped due to: The exhaust flowrate of RCIC is approximately equal to that of decay heat, and is thus consistent with the basis used for determining the Primary Containment Pressure Limit and Elevated Torus pressure will cause the RCIC turbine to trip before the HPCI turbine would trip. (Refer to the discussion of Caution 4)

D. Incorrect - (1) The RCIC turbine exhaust is at the approximate same level, but the RCIC is not tripped due to: The exhaust flowrate of RCIC is approximately equal to that of decay heat, and is thus consistent with the basis used for determining the Primary Containment Pressure Limit and Elevated Torus pressure will cause the RCIC turbine to trip before the HPCI turbine would trip. (Refer to the discussion of Caution 4).

Technical Reference(s): EOP 2 Bases, page 13 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments6-10-11-NRC OK enhanced 6-14-11-Old distractors C & D could be argued, changed to RCIC turbine.

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295021 AA2.03 Importance Rating 3.5 Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING: Reactor water level Question: RO Question # 52 The plant is in Mode 4 with RHR "A" in shutdown cooling with the following conditions:

• RPV water level momentarily drops to 168 inches and is recovered to 173 inches What is the effect on Shutdown Cooling?

A. Shutdown Cooling remains in service.

B. The A RHR pump trips directly due to RPV level. The inboard and outboard Shutdown Cooling Isolation valves go CLOSED.

C. The A RHR pump remains in service but only on minimum flow. The inboard and outboard Shutdown Cooling Isolation valves go CLOSED.

D. The A RHR pump trips because a loss of suction path is sensed by the pump trip circuitry when the Shutdown Cooling Isolation valves begin to CLOSE.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The pump tripped and the valves closed B. Incorrect - The pump tripped due to loss of suction path NOT low RPV level C. Incorrect - The pump tripped and the valves closed D. Correct - The valves close at 170 RPV level. When they begin to close (not fully open) the pump trips because a loss of suction path is sensed by the pump trip circuitry.

SD 149 Rev.11. pages 11, 32, Technical Reference(s): (Attach if not previously provided)

Figure 2 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTS 10960 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3-11-revised distractor D 6-10-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295038 EA2.04 Importance Rating 4.1 Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Source of off-site release Question: RO Question # 53 The plant is operating in MODE 1 at 100% power with the following conditions:

• Annunciator 1C03A A-4, OFFGAS VENT PIPE RM-4116A/B HI-HI RAD alarms

• Standby Gas Treatment System initiates Which of the following choices below could be the source for the above alarm?

(1) A Reactor Recirc pump seal leak (2) A Condenser Bay steam leak (3) A RWCU Pump seal leak (4) A leak in the Torus Room A. (1), (2) and (3)

B. (2), (3) and (4)

C. (1), (3) and (4)

D. (1), (2) and (4)

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - (1) would be contained in the drywell B. Correct - See SD 733 Figures 4,5,6 C. Incorrect - (1) would be contained in the drywell D. Incorrect - (1) would be contained in the drywell Technical Reference(s): SD 733 Figures 4,5,6 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 11 55.43 Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295031 EK1.03 Importance Rating 3.7 Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL: Water level effects on reactor power Question: RO Question # 54 During execution of ATWS-RPV Control, it is required to lower RPV Water Level to at least 87 inches.

Which of the following describes the reason for this requirement?

It is required to lower RPV Water Level to at least 87 inches to _______.

A. reduce natural circulation and limit the peak power level to below the fuel thermal limits B. uncover the feedwater spargers to reduce subcooling and limit the onset of reactor power / core flow instabilities C. isolate RWCU to prevent boron removal by the system and limit the peak power level to below the fuel thermal limits D. trip the operating Recirculation Pumps to reduce forced circulation and limit the onset of reactor power / core flow instabilities Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - the concern is core instabilities.

B. Correct - IAW EOP ATWS Bases Continuous Recheck Statement - The conditions expressed in this Continuous Recheck Statement, combined with the inability to shutdown the reactor through control rod insertion, dictate a need to promptly reduce reactor power in order to prevent or mitigate the consequences of any large irregular neutron flux oscillations induced by neutronic/thermal-hydraulic instabilities. This is accomplished by transferring to entry point 7 and lowering RPV water level to +87 inches in Step /L-2. An RPV water level of +87 inches is 2 feet below the lowest nozzle in the feedwater sparger. This places the feedwater spargers in the steam space providing effective heating of the relatively cold feedwater and eliminating the potential for high core inlet subcooling.

C. Incorrect - RWCU is verified isolated, but the reason for lowering level to 87 inches is NOT based on RWCU automatic isolation at 119.5 inches D. Incorrect - RR Pumps will be verified tripped if power is above 5%, but the reason for lowering level to 87 inches is NOT based on RR Pump ATWS RPT at 119.5 inches.

EOP ATWS Bases Rev 14 page Technical Reference(s): (Attach if not previously provided) 15 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTS 11294 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6-3-11-revised A to remove uncover fuel and just stated reduce natural circulation and 6-10-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295026 2.1.28 Importance Rating 4.1 Conduct of Operations: Knowledge of the purpose and function of major system components and controls. (Suppression Pool High Water Temp).

Question: RO Question # 55 A transient resulted in the following plant conditions:

• RPV level is 60 inches and steady

• RPV pressure is 800 psig and lowering slowly

• Torus and Containment Sprays have been initiated once

• Drywell Pressure is 1.6 psig and steady

• Drywell Temperature is 100°F and steady

• Torus Temperature is 102°F rising slowly The Control Room Supervisor directs the operator to maximize torus cooling. Is this allowed by current plant conditions? Why or why not?

A. Yes, since adequate core cooling has been assured, the operator may establish Torus Cooling.

B. Yes, since there is less than a 2 psig drywell pressure signal, the operator may establish Torus Cooling.

C. No, since RPV level is less than 64" and drywell pressure is less than 2 psig, Torus Cooling may NOT be established.

D. No, since RPV pressure is 800 psig and LPCI loop select has selected a loop, Torus Cooling may NOT be established.

Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - There is precaution on verifying adequate core cooling and with 60 in the RPV adequate core cooling is assured, however the torus cooling valves cannot be opened with less than 2 psig in the drywell and the LPCI signal still in.

B. Incorrect - The torus cooling valves cannot be opened with less than 2 psig in the drywell and the LPCI signal still in.

C. Correct - IA OI-149, Sect 5.3, pg 32, The Containment Spray and Cooling valves are interlocked closed when Drywell pressure is < 2 psig with a LPCI Initiation signal present. The LPCI signal is still present because the RPV water level is <119.5 inches.

D. Incorrect - Torus cooling could still be placed in service with these conditions IF DW pressure was >2psig.

Technical Reference(s): OI-149, pg 32 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # 19019 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295004 2.4.31 Importance Rating 4.2 Emergency Procedures / Plan: Knowledge of annunciator alarms, indications, or response procedures. (Partial or Total Loss of DC Pwr)

Question: RO Question # 56 The plant is operating in MODE 1 at 100% power when the following alarm occurs:

• 1C08A C-8, INSTRUMENT AC 1Y11 UNDERVOLTAGE OR INVERTER TROUBLE What is the plant response to this annunciator?

If the alarm was caused by a ____.

A. low inverter AC OUTPUT, the Reactor Water Cleanup system will isolate.

B. low inverter AC OUTPUT, the Reactor Water Cleanup pumps will trip but the system will NOT isolate.

C. low voltage condition on Instrument Bus 1Y11, the A Recirc Pump will trip.

D. low voltage condition on Instrument Bus 1Y11, the A Recirc Pump scoop tube will lock up.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Low AC output results only in a trouble lamp on 1D15 B. Incorrect - Low AC output results only in a trouble lamp on 1D15 C. Incorrect - The pump does not trip but the scoop tube locks up D. Correct - IAW ARP 1C08A C-8, Section 2.2, If the cause was due to a low voltage condition on the bus - RWCU Pumps 1P-205A and B trip, RWCU System isolates and Recirc Pump 1P-201A scoop tube locks up As Is.

Technical Reference(s): 1C08A C-8 Sections 1 and 2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3-11-spelled out RWCU 6-9-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295016 2.1.2 Importance Rating 4.1 Conduct of Operations: Knowledge of operator responsibilities during all modes of plant operation.

Question: RO Question # 57 The plant was operating in MODE 1 at 100% power when a NON-FIRE event occurred that required evacuation of the Control Room per AOP-915, Shutdown Outside the Control Room.

The following actions have been completed:

• Manual SCRAM has been inserted.

• ALL RODS have been verified inserted using the "One Rod Permissive" technique.

• The 1C05 operator has completed the "as time permits" actions of AOP-915 and evacuated the Control Room.

When the 1C05 Operator left the control room the Mode Switch would be in_____.

A. RUN B. REFUEL C. SHUTDOWN D. START & HOT STBY Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct -AOP 915 requires Reactor Mode Switch placed in RUN following Reactor Scram actions.

B. Incorrect - REFUEL position was used to verify ALL RODS IN.

C. Incorrect - SHUTDOWN is the normal post-scram Mode Switch position.

D. Incorrect - START & HOT STBY may be selected if the candidate knows a position other than SHUTDOWN is used, but doesnt know the correct position.

Technical Reference(s): AOP 915 Rev.41, Step 4.0 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None 94.28.01.03 Learning Objective: (As available)

Question Source: Bank # WTS Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295028 EK1.02 Importance Rating 2.9 Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE: Equipment environmental qualification Question: RO Question # 58 In accident conditions, IAW EOP-2, Primary Containment Control, action is required if drywell temperature cannot be restored and maintained below 280°F.

Why is action required at this temperature?

A. At this temperature, closure of the MSIVs, if required, could not be assured because the MSIV Solenoids have reached their environmental qualification temperature limit.

B. Implementation of Drywell Spray above this temperature will NOT prevent exceeding the drywell analytical withstand temperature.

C. To provide margin to the temperature where the ADS SRVs and ADS Solenoids may not function if required to depressurize to RPV.

D. Torus to Drywell Vacuum Breakers are not designed to operate at this temperature and may not be able to function and minimize a Torus pressure spike under LOCA conditions.

Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The MSIVS and their solenoids are not a concern at this point in the EOPs.

They are in all probability already closed due to a LOCA condition.

B. Incorrect - Drywell Spray if not already initiated may prevent exceeding the drywell analytical withstand temperature however the EOPs require an ED in this case for that purpose C. Correct - IAW EOP-2 Bases - The EQ rating of equipment in the drywell, specifically the ADS valves and ADS solenoids, is 340 °F for a significant time. Although EQ analysis indicates that the ADS valves are operable for an extended period of time at 340 °F, management expectation is that operators will direct ED before 340 °F to ensure that the EQ limits and the drywell analytical withstand temperature is not exceeded.

D. Incorrect - the design temperature of the Drywell is 281F Technical Reference(s): EOP-2 Bases Rev.13 page 41 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 5 55.43 Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons for these operating characteristics.

Comments:

6-3-1-revised stem 6-10-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 295035 EK1.02 Importance Rating 3.7 Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Radiation release Question: RO Question # 59 The plant is in MODE 5 when a fuel handling accident occurs with the following conditions:

• No OPDRVs are in progress

• No PCIS Group III isolation setpoints have been exceeded during the event

• The A Standby Gas Treatment System is manually initiated with isolation IAW OI-170, Standby Gas Treatment System

• Secondary Containment Isolation Damper 1V-AD-19A fails to close What is the operational implication of this condition?

Possible ____ .

A. entry into LCO 3.0.3 due to loss of Secondary Containment B. release via Reactor Building Exhaust Fans 1VEF11A or 1VEF11B C. excessive flow thru the operating SBGT train D. unfiltered release from the Secondary Containment Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - loss of Secondary Containment is not an LCO 3.0.3 issue B. Incorrect - A Group 3 isolation signal will trip the 11A & 11B fans C. Incorrect - SBGT have flow controllers to control the flow going thru the SBGT train D. Correct - With only one division of the Group 3 in, and one isolation damper failed to close, there is a possibility of unfiltered release from the Secondary Containment thru the open isolation damper.

Technical Reference(s): SD 733 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTS 11401 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-10-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 500000 EK2.03 Importance Rating 3.3 Knowledge of the interrelations between HIGH CONTAINMENT HYDROGEN CONCENTRATIONS the following: Containment Atmosphere Control System Question: RO Question # 60 Which one of the following describes how primary containment Oxygen and Hydrogen concentrations are monitored?

(1) O2 Concentration (2) H2 Concentration A. (1) Is continuously monitored during normal and emergency operations (2) Can be monitored under accident conditions ONLY B. (1) Is continuously monitored during normal and emergency operations (2) Is continuously monitored during normal and emergency operations C. (1) Is continuously monitored during normal and emergency operations (2) Can be monitored during normal and emergency operations D. (1) Can be monitored under accident conditions ONLY (2) Can be monitored under accident conditions ONLY C

Proposed Answer:

ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - O2 is normally monitored. H2 is NOT normally monitored. Both may be monitored under emergency conditions.

B. Incorrect - O2 is normally monitored. H2 is NOT normally monitored. Both may be monitored under emergency conditions.

C. Correct - O2 is normally monitored. H2 is NOT normally monitored. Both may be monitored under emergency conditions.

D. Incorrect - O2 is normally monitored. H2 is NOT normally monitored. Both may be monitored under emergency conditions..

Technical Reference(s): SD 573 Rev.10 page 34 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-10-11-NRC OK with revision - still unsat ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 295007 AK3.05 Importance Rating 3.0 Knowledge of the reasons for the following responses as they apply to HIGH REACTOR PRESSURE: Low pressure system isolation Question: RO Question # 61 The plant is in MODE 3, Shutdown Cooling is in service with A RHR Pump in service

• Reactor coolant temperature and pressure are slowly rising.

• RPV level is 190 inches stable, maintaining on dump flow The Shutdown Cooling automatic isolation actions have all occurred as designed.

The reason for these automatic actions is to prevent ____.

A. RHR suction piping overpressurization B. steam voiding in the RHR pump seals C. overpressurizing the RHR pump seals D. establishing a drain path from the RPV to the torus Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - The Reactor Steam Dome Pressure High Function is provided to isolate the shutdown cooling portion of the Residual Heat Removal (RHR) System (i.e., the shutdown cooling suction valves). This interlock is provided only for equipment protection to prevent an intersystem LOCA scenario (i.e., a break of the low pressure RHR suction piping caused by exposure to relatively high pressure RPV fluid)

B. Incorrect - this would not be a primary concern C. Incorrect - overpressurizing the piping is the concern D. Incorrect - there are valve interlocks that prevent this from occurring.

Technical Reference(s): TA Bases 3.3.6.1 6.a. (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTS 10569 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-10-11-NRC OK- changed to F ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 295013 AA1.01 Importance 3.9 Rating Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE: Suppression pool cooling Question: RO Question # 62 A loss of coolant accident has occurred. The following plant conditions exist:

• Reactor Water Level +110 inches and slowly rising

• Drywell Pressure is 2.5 psig and slowly lowering

• Torus Temperature is 110 degrees F. and slowly rising

• The Essential Buses are being powered from the Standby Transformer

• A & B ESW pumps are in service

• A, B and C RHR pumps are in service

• A, B, and D RHRSW pumps are in service Prior to placing the D RHR pump in Torus Cooling, which one of the following describes whether HS-1903C-Enable Containment Spray Valves, must be placed in the MAN position and if a running RHR or RHRSW pump must be removed from service IAW OI 149 QRC 2.

A. HS-1903C must be placed in MAN and the B RHR pump OR the B RHRSW pump OR the D RHRSW pump must removed from service.

B. HS-1903C is NOT required to be placed in MAN and the B RHR pump OR the B RHRSW pump OR the D RHRSW pump must removed from service.

C. HS-1903C must be placed in MAN and no RHR or RHRSW pumps are required to be removed from service.

D. HS-1903C is NOT required to be placed in MAN and and no RHR or RHRSW pumps are required to be removed from service..

Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - IAW OI 149 QRC 2 - CAUTION While the Essential buses are powered from the Standby Transformer, do not run more than a total combination of 3 RHR/RHRSW pumps on each essential bus. (e.g. 2 RHR pumps & 1 RHRSW pump , or 1 RHR pump

& 2 RHRSW pumps). With a combination of 3 RHR/RHRSW pumps in service, stop one pump before starting the out of service pump.

If a LPCI HI Drywell pressure condition (2 # ) exists, place HS-2001C[1903C] Enable Containment Spray Valves in the MAN position.

B. Incorrect - The Enable Containment Spray Valves HS must be placed in manual C. Incorrect - one of the listed pumps must first be removed from service D. Incorrect -the Enable Containment Spray Valves HS must be in the MAN position and one of the listed pumps must first be removed from service Technical Reference(s): OI 149 QRC 2 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 7 55.43 Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Comments:

6-3-11-revised question 6-10-11-NRC OK with change- enhanced not unsat ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 295036 EA2.03 Importance Rating 3.4 Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT HIGH SUMP/AREA WATER LEVEL: Cause of the high water level Question: RO Question # 63 The plant is operating in MODE 1 at 100% power with the following conditions:

• A large leak in the Drywell from the RBCCW System occurs

• A fast power reduction is performed IAW IPOI-4, Shutdown

• The reactor is manually scrammed

• Drywell and Reactor Building Sump High Sump Level alarms are IN

• All scram signals are clear and the scram is reset Assuming no other operator actions have been taken, which of the following is correct concerning these conditions?

A. The Reactor Building Equipment Drain Sump is filling from the Scram Discharge Volume header and pumps will transfer water to Radwaste with no further operator action.

B. The Reactor Building Floor Drain Sump is filling from the Scram Discharge Volume header and pumping down to the Floor Drain Collector Tank.

C. The Drywell Equipment Drain Sump is filling from the RBCCW leak and pumps will transfer water to Radwaste with no further operator action.

D. The Drywell Floor Drain Sump is filling from the RBCCW leak and pumping down to the Floor Drain Collector Tank.

Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - IAW SD 920-1, the CRD Hydraulic system drains to the reactor building equipment drain sump. When the scram is reset, the SDV will drain to that sump and pump to the radwaste collector tank.

B. Incorrect - The SDV does not drain into the floor drain C. Incorrect - The Drywell Equipment drain would be isolated and not pumping down until PCIS Isolation signal was clear and reset.

D. Incorrect - The Drywell Floor drain would be isolated and not pumping down until the PCIS group 2 signal was clear and reset.

SD 920-1 Rev.4, page 18, figures Technical Reference(s): (Attach if not previously provided) 1,2,5,6 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 4 55.43 Secondary coolant and auxiliary systems that affect the facility.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 295022 2.1.20 Importance Rating 4.6 Conduct of Operations: Ability to interpret and execute procedure steps. (Loss of CRD pumps)

Question: RO Question # 64 The plant is operating in MODE 1 at 100% power with the following conditions:

• The A CRD pump out of service to replace the motor bearings

• The 1A4 bus suffers a lockout trip and is de-energized Due to a loss of drywell cooling, the CRS directs a manual reactor scram.

What will be the effect on the control rods and subsequent actions?

A. Control rods will fully insert slower than normal on the scram. IPOI-5 and EOP-1 will be entered.

B. Control rods will fully insert at normal speed on the scram. IPOI-5 and EOP-1 will be entered..

C. Control rods will NOT fully insert on the scram. EOP-1 will be entered and transferred to EOP-ATWS for actions to be directed. Actions directed will be for a LOW power ATWS.

D. Control rods will fully insert on the scram. EOP-1 will be entered and then IPOI-5. A CRD pump must be re-started before the scram is able to be reset.

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - Control rods will insert into core without CRD pump running. The scram time testing STP has the charging water isolation valve shut prior to running the scram time test. This is equivalent to losing both CRD pumps. The rods do not insert slower than with the CRD pumps running.

B. Correct - Control rods insert without CRD pump, EOP 1 will be required to be entered on the RPV level shrink, and IPOI-5 is the scram procedure.

C. Incorrect - Control rods will insert into core without CRD pump running. This answer is plausible if the candidate believes that the rods will partially insert, but not go full in.

D. Incorrect - CRD pump is not required to reset the scram.

IPOI-5 (reset scram section)

Technical Reference(s): SD 255 (ball check valve (Attach if not previously provided) discussion)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # DAEC 19984 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 6 55.43 Design, components, and functions of reactivity control mechanisms and instrumentation.

Comments:

6-10-11-NRC OK with changes ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 295008 AA1.03 Importance Rating 3.3 Ability to operate and/or monitor the following as they apply to HIGH REACTOR WATER LEVEL: Main steam system Question: RO Question # 65 When carrying out RPV FLOODING EOP with 62 control rods not full in, what is the required position of the Main Steam Isolation Valves (MSIVs), and what is the reason for that requirement?

Main Steam Isolation Valves are required to be ____.

A. open, to allow Main Steam flow to assist in rapidly depressurizing the RPV and ensure boron is mixed throughout the vessel.

B. open, to allow flooded RPV indications to be obtained from Main Steam Line Flow Instruments.

C. shut, the primary concern is to avoid excessive water inventory loss from the RPV during flooding.

D. shut, to ensure adequate boron concentration in the vessel and avoid damage to downstream equipment.

Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The MSIVs are shut in EOP-ATWS step RPV/F-12. Boron would be diluted if the MSIVs were open B. Incorrect - The MSIVs are shut per the EOP C. Incorrect - Inventory loss is not the concern.

D. Correct - The MSIVs are shut in EOP-ATWS step RPV/F-12. IAW the bases, If the MSIVs were not closed, boron would be lost from the RPV when water level reached the elevation of the main steam lines. Leaving the MSIVs open would also risk damage to downstream equipment that might be needed during later recovery actions.

EOP-ATWS Technical Reference(s): EOP-ATWS Bases Rev 12 page (Attach if not previously provided) 23 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 2 55.43 General design features of the core, including core structure, fuel elements, control rods, core instrumentation, and coolant flow.

Comments:

6-3-11-changed ONLY to primary in distractor C 6-10-11-NRC OK with change ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G1 2.1.1 Importance Rating 3.8 Conduct of Operations: Knowledge of conduct of operations requirements Question: RO Question # 66 An ANSOE is on watch with an ILT student. The Shift Technical Advisor (STA) is NOT an SRO. A Journeyman I&C Tech performing an STP requests the ANSOE bypass the "A" APRM for his STP. The Reactor Engineer is in the Control Room to talk with the CRS.

Which personnel may serve as the Peer Check for the ANSOE as the "A" APRM is bypassed?

A. STA: May NOT Peer Check Reactor Engineer: May Peer Check B. STA: May Peer Check Reactor Engineer: May Peer Check C. STA: May NOT Peer Check Reactor Engineer: May NOT Peer Check D. STA: May Peer Check Reactor Engineer: May NOT Peer Check Proposed Answer: D ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - STA may provide Peer Check, RE may not Peer Check based upon not being familiar with the task.

B. Incorrect - Peer Checker quals should be consistent with that of the performer. An STA is allowed to peer check in the Control Room, while the RE would not be familiar with the task.

C. Incorrect - STA may provide Peer Check.

D. Correct - STA may provide Peer Check. Reactor Engineer may NOT provide Peer Check.

OP-AA-100-1000 Technical Reference(s): (Attach if not previously provided)

PI-AA-103-1000 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-3-11-revised with bank question 6-10-11-NRC OK with change ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G1 2.1.26 Importance Rating 3.4 Conduct of Operations: Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

Question: RO Question # 67 In accordance with OP-AA-101, Clearance and Tagging, which one of the following conditions would require double valve protection?

Any system where the isolated portion of the system contains A. conditions equal to or greater than 200 psig or 500°F.

B. conditions equal to or greater than 500 psig or 200°F.

C. radioactive concentrations in excess of 10CFR20 Appendix C limits and/or temperatures equal to or greater than 212°F D. radioactive concentrations in excess of 10CFR20 Appendix E limits and/or temperatures equal to or greater than 212°F.

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The values are greater than 500 psig or 200°F.

B. Correct - When isolating high energy systems (>500 psi or >200°F on piping >3/8" diameter) or hazardous chemical systems (as determined by the Safety Department or indicated in the MSDS information), then double valve isolation SHALL be used (two valves in series) when available or practical.

C. Incorrect - The values are greater than 200°F and there are no restrictions based on radiation.

D. Incorrect - The values are greater than 200°F and there are no restrictions based on radiation.

Technical Reference(s): OP-AA-101, Att 6, pg 94 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # G2 2.2.39 Importance Rating 3.9 Equipment Control: Knowledge of less than or equal to one hour technical specification action statements for systems.

Question: RO Question # 68 The plant is operating in MODE 1 at 100% power when the A Recirculation MG set trips due to an electrical fault. Due to an operator error, the RO closes the B Recirculation Pump Suction Valve instead of the A Recirculation Pump Discharge Valve.

What action must be taken?

A. Take action to insert all insertable control rods within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> B. Immediately scram the reactor and carry out IPOI 5 C. Enter LCO 3.0.3 immediately and be in MODE 2 within 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> D. Enter AOP 255.2, Power/Reactivity Abnormal Change, and insert control rods per the current rod pull sheet.

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - The shutting of the only operating RR pump suction valve will trip that RR pump. This leaves the reactor in a natural circulation mode, which is prohibited by Tech Specs., and requires an immediate scram.

B. Correct - The shutting of the only operating RR pump suction valve will trip that RR pump. This leaves the reactor in a natural circulation mode, which is prohibited by Tech Specs., and requires an immediate scram.

C. Incorrect - The shutting of the only operating RR pump suction valve will trip that RR pump. This leaves the reactor in a natural circulation mode, which is prohibited by Tech Specs., and requires an immediate scram.

D. Incorrect - Entry to AOP is required however, The shutting of the only operating RR pump suction valve will trip that RR pump. This leaves the reactor in a natural circulation mode, which is prohibited by Tech Specs., and requires an immediate scram.

Technical Reference(s): SD 264 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-3-11-revised distractor D 6-10-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 2 K/A # G2 2.2.14 Importance Rating 3.9 Equipment Control: Knowledge of the process for controlling equipment configuration or status.

Question: RO Question # 69 Which one of the following are approved methods of deviating from the Locked Valve List?

1. Component clearance

2. An approved procedure

3. Work Control Supervisor direction

4. Operations Shift Manager direction A. 1, 2, 4 B. 1, 3, 4 C. 2, 3, 4 D. 1, 2, 3 Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - Locked valves may only be manipulated from their required position under procedures that control the testing or operation of plant systems that are prepared and approved per site administrative control procedures. Examples include an OI, RFP, RWH, SPTP, or MAT. A Clearance can direct the change of position, as well as the OSM direction under emergency direction.

B. Incorrect - WCCS cannot direct the deviation from the Lock Valve List.

C. Incorrect - WCCS cannot direct the deviation from the Lock Valve List.

D. Incorrect - WCCS cannot direct the deviation from the Lock Valve List.

Technical Reference(s): ACP-1410.9, pg 3 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # DAEC #20496 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-10-11-changed question.

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # G3 2.3.7 Importance Rating 3.5 Radiation Control: Ability to comply with radiation work permit requirements during normal or abnormal conditions.

Question: RO Question # 70 The #1 Traversing In-Core Probe (TIP) detector is stuck in the core, all other TIP detectors are in their shields. An Operator and Health Physics Technician must enter the TIP Room to verify the position of the TIP takeup reel.

In accordance with OI-878.6, Traversing In-Core Probe System, and HPP 3104.01, Control of Access to High Radiation Areas and Above, which one of the following is required?

Prior to entry into the TIP Shield area the ...

A. TIP machines shall be tagged out and the Operations Manager must sign on the tagout.

B. TIP machines shall be tagged out and the Health Physics Supervisor or designee must sign on the tagout.

C. Health Physics Supervisor shall discuss the work plans and exposure control plans with the CRS and Operator.

D. CRS shall discuss the work plans and exposure control plans with the Health Physics Technician and Operator.

Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - There is no requirement for the Ops Suprv to sign on the tagout.

B. Correct - In accordance with HPP 3104.01 Control of Access to High Radiation Areas and Above, entries into the TIP Shield area and/or for entries to work on the TIP machine that would have the potential to draw the TIP into the TIP machine, the TIP machines shall be tagged out and the Health Physics Supervisor or designee shall be required to sign on the tagout.

C. Incorrect - A briefing is required if the TIP can NOT be tagged out. When the work to be performed prevents the machines from being tagged out, the Health Physics Technician providing coverage for work in the area will discuss work plans and exposure control plans with the CRS and Health Physics Supervisor.

D. Incorrect - A briefing is required if the TIP can NOT be tagged out. When the work to be performed prevents the machines from being tagged out, the Health Physics Technician providing coverage for work in the area will discuss work plans and exposure control plans with the CRS and Health Physics Supervisor.

Technical Reference(s): OI-878.6, pg 4 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 3 K/A # G3 2.3.4 Importance Rating 3.2 Radiation Control: Knowledge of radiation exposure limits under normal or emergency conditions.

Question: RO Question # 71 ACP-1411.25, Planned Special Exposures permits a worker who has critical skills and that is necessary for a particular job can be authorized to receive an exposure in ADDITION to the routine occupational exposure limit.

The workers Annual (TEDE) Exposure Limited can be raised to ___(1)___ if authorized by the

___(2)___.

A. (1) 5 Rem (2) Plant Manager, Nuclear B. (1) 10 Rem (2) Plant Manager, Nuclear C. (1) 5 Rem (2) Manager, Radiation Protection D. (1) 10 Rem (2) Manager, Radiation Protection Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - The individual(s) receiving a PSE are limited to the following dose from all PSEs in one year, 5 Rems TEDE. The Plant Manager, Nuclear is responsible for the authorization of a PSE B. Incorrect - The individual(s) receiving a PSE are limited to the following dose from all PSEs in one year, 5 Rems TEDE.

C. Incorrect - The Plant Manager, Nuclear is responsible for the authorization of a PSE D. Incorrect - The individual(s) receiving a PSE are limited to the following dose from all PSEs in one year, 5 Rems TEDE. The Plant Manager, Nuclear is responsible for the authorization of a PSE Technical Reference(s): ACP-1411.25, pgs 4 & 5 (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G4 2.4.28 Importance Rating 3.2 Emergency Procedures / Plan: Knowledge of procedures relating to a security event (non-safeguards information).

Question: RO Question # 72 The plant is in normal full power operation..

The NRC has just called on the ENS phone to inform the DAEC of a confirmed terrorist attack with an explosives filled aircraft at the Brunswick plant in North Carolina.

The FAA has grounded all aircraft nationally. However, they are watching two small planes headed towards the Cedar Rapids area from the North West that have not yet responded to radio communications. Time to the site is 40 minutes.

In accordance with AOP 914 Security Events which operator actions if any are appropriate at this time?

A. Reduce core flow, manually scram the reactor, and evacuate the site.

B. Commence a rapid downpower of the reactor using IPOI 4, Fast Power Reduction.

C. Remain at full power, back out of any STPs that are in progress and verify all ECCS operable.

D. Remain at full power, increase plant monitoring, and take NO further actions until a plane is within 30 minutes of the site.

Proposed Answer: C ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - This action would be correct for a Airborne Attack Probable (in the next 30 minutes).

B. Incorrect - Per Tab 3, the plant may remain at full power.

C. Correct - The event described is an Attack on US Soil and meets the definition of an "informational airborne attack. Actions are from Tab 3.

D. Incorrect - Per AOP 914, the plant may remain at full power however many preliminary actions must be taken, including backing out of any STPs that are in progress and verify all ECCS operable.

AOP 914, Tab 3, pg 22 Technical Reference(s): (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # DAEC #10044 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-3-11-leave as is 6-10-11-NRC OK with change ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G4 2.4.16 Importance Rating 3.5 Emergency Procedures / Plan: Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, AOP's and SAMG's.

Question: RO Question # 73 The plant is operating in MODE 1 at 93% power with the following conditions:

• The Torus developed an unisolable leak

• EOP-1, RPV Control, was entered after the scram due to RPV level shrink

• RPV level has since been restored to 190 inches

• RPV pressure is 920 psig and being controlled by EHC Pressure Set The CRS directs the RO to perform SEP 307, Rapid Depressurization with Bypass Valves, to anticipate Emergency Depressurization due to Torus Level continuing to decrease uncontrollably.

(1) Is this an appropriate action at this time?

Assume the SEP 307 actions were NOT taken as above, when the CRS directs Emergency Depressurization for this event, only 1 SRV would open. The CRS then directs the BOP to perform SEP 307 as an Alternate Depressurization System.

(2) Is this an appropriate action at this time?

A. (1) Rapid Depressurization with Bypass Valves is appropriate (2) performance of SEP 307 as an Alternate Depressurization System is appropriate B. (1) Rapid Depressurization with Bypass Valves is appropriate (2) performance of SEP 307 as an Alternate Depressurization System is NOT appropriate C. (1) Rapid Depressurization with Bypass Valves is NOT appropriate (2) performance of SEP 307 as an Alternate Depressurization System is appropriate D. (1) Rapid Depressurization with Bypass Valves is NOT appropriate (2) performance of SEP 307 as an Alternate Depressurization System is NOT appropriate Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - SEP 307 Purpose identifies its use for when ED is anticipated and for when less than the minimum number of SRVs has opened during ED. This SEP may not be used to anticipate ED during ALC or ATWS transients, so there are times when it would not be appropriate B. Incorrect - Listed as a Table 8 Alternate Depressurization System. As long as the MSIVs remain open, this SEP is appropriate. Selected if it is believed that all alternate systems go to the Torus C. Incorrect - SEP would not be appropriate before ED for two other types of transients, but would be for this one D. Incorrect - SEP would not be appropriate before ED for two other types of transients, but would be for this one. Listed as a Table 8 Alternate Depressurization System. As long as the MSIVs remain open, this SEP is appropriate. Selected if it is believed that all alternate systems go to the Torus SEP 307 Technical Reference(s): (Attach if not previously provided)

EOP Bases, EOP-1 Page 34 Proposed References to be provided to applicants during examination: None 96.06.06.06 Learning Objective: (As available) 95.00.00.20 Question Source: Bank # 2005 NRC #74 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2005 Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility.

Comments:

6-10-11-NRC OK with change ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 4 K/A # G4 2.4.8 Importance Rating 3.8 Emergency Procedures / Plan: Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Question: RO Question # 74 The plant is operating in MODE 1 at 100% power with the following conditions:

• A loss of Startup Transformer 1X3 and Aux Transformer 1X2 occurred

• The reactor automatically scrammed

• EOP 1, RPV Control, was entered due to RPV level shrink on the scram

• IPOI 5, Reactor Scram, has been entered

• AOP 304.1, Loss of 4160 VAC Non Essential Power, has been entered Two minutes later Torus Water Temperature is 95°F and rising.

Which of the following actions is required?

A. Concurrently enter EOP-2, PRIMARY CONTAINMENT CONTROL B. Continue IPOI-5, REACTOR SCRAM and monitor Torus Water Temperature, entry into EOP 2 is not required C. Exit BOTH IPOI-5 and AOP-304.1 and enter EOP-2, PRIMARY CONTAINMENT CONTROL D. Exit IPOI-5, REACTOR SCRAM, and enter EOP-2, PRIMARY CONTAINMENT CONTROL Proposed Answer: A ILT Exam 7/12/2011

Explanation (Optional):

A. Correct - with Torus Water Temperature above 95F, it is required to concurrently enter EOP-2, Primary Containment Control B. Incorrect - would be true below 95F Torus Water Temperature C. Incorrect - would be true after actions of BOTH IPOI-5 and AOP-304.1 are complete D. Incorrect - would be true if IPOI-5 actions were complete prior to exceeding 95F Torus Water Temperature Technical Reference(s): EOP-2 entry condition (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # WTS 11260 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 10 55.43 Administrative, normal, abnormal, and emergency operating procedures for the facility Comments 4-4 is there a procedure usage reference we could use?

6-3 need to provide learning objective 6-10-11-NRC OK ILT Exam 7/12/2011

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group # 1 K/A # G1 2.1.40 Importance Rating 2.8 Conduct of Operations: Knowledge of refueling administrative requirements Question: RO Question # 75 The plant was shutdown fourteen days ago for a refueling outage, with maintenance occurring that has the potential to drain the reactor vessel (OPDRV).

An Operator contacts the Control Room and informs you that someone has blocked open both reactor building airlock doors.

Which one of the following actions is required?

A. Within four hours verify one airlock door closed or stop maintenance with the potential to drain the reactor vessel (OPDRV)

B. Immediately stop maintenance with the potential to drain the reactor vessel (OPDRV) while initiating action to close at least one air lock door C. Within four hours verify one airlock door closed or stop any refueling activities on the Refuel Floor, maintenance with the potential to drain the reactor vessel (OPDRV) may continue D. Immediately stop any refueling activities on the Refuel Floor and initiate action to close at least one air lock door, maintenance with the potential to drain the reactor vessel (OPDRV) may continue Proposed Answer: B ILT Exam 7/12/2011

Explanation (Optional):

A. Incorrect - immediate actions is required by T.S.

B. Correct - With Secondary Containment inoperable initiate actions to suspend OPDRVs.

C. Incorrect - Immediately and maintenance with the potential to drain the reactor vessel (OPDRV) must be stopped D. Incorrect - Immediately and maintenance with the potential to drain the reactor vessel (OPDRV) must be stopped Technical Reference(s): T.S. 3.6.4.1.C (Attach if not previously provided)

Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank # Hope Creek Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

6-4 this is RO knowledge because this question is based on information that is above the ACTION line in Tech. Specs.

6-10-11 NRC OK ILT Exam 7/12/2011