ML103540549

FC-2010-09-DRAFT Written Exam - Delay Release Date

ML103540549
Person / Time
Site:	Fort Calhoun
Issue date:	09/17/2010
From:	Brian Larson Division of Reactor Safety IV
To:
References
50285/10-301
Download: ML103540549 (123)

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CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 001 The plant is in hot shutdown following a loss of offsite power. The following plant conditions exist:

x 4160 volt buses 1A3 and 1A4 are both energized by the Diesel-Generators x LRC-101X indicates 46% with a slowly rising trend x LRC-101Y (controlling channel) indicates 45% with a slowly rising trend x PRC-103X indicates 1910 psia and lowering x PRC-103Y (controlling channel) indicates 1912 psia and lowering x Charging flow indicates 40 gpm x Letdown flow indicates 32 gpm x One charging pump is running x Pressurizer Quench Tank level and pressure are steady x Reactor Vessel Level indicates 100%

x Representative CET temperature indicates 575°F Which one of the following events could cause these indications?

A. Pressurizer Spray Valve, PCV-103-1, inadvertently opened.

B. Auxiliary Spray Valve, HCV-249, inadvertently opened.

C. Pilot Operated Relief Valve, PCV-102-1, inadvertently opened.

D. RCS Void Formation is occurring.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 1 Revision: 0 KA #: 000008 AA2.19 Tier 1 Group 1 Pressurizer Vapor Space Accident Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident:

PZR spray valve failure, using plant parameters Importance: 3.4 / 3.6 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

PLOT and PREDICT the following parameters for the transients listed in objective 1.2:

EXPLANATION:

Choice B is correct, with the aux spray valve open, charging flow will go though aux spray valve causing a pressure decrease. Choice A is incorrect because there will be no main spray low during a loss of offsite power with no reactor coolant pumps running. C is incorrect because pressurizer quench tank level and pressure are steady. D is incorrect because adequate subcooling exists and RV level is 100%

KA#: 000008 AA2.19 Bank Ref #: 07-17-33 002 LP# / Objective: 0715-12 01.03 Exam Level: RO-5 Cognitive Level: HIGH Source: MODIFIED

Reference:

STM 12 Handout: STEAM TABLES CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 002 The following plant conditions exist following an automatic reactor trip and Pressurizer Pressure Low Setpoint (PPLS) actuation:

x Pressurizer Pressure is 725 psia and lowering x Pressurizer level is 0%

x Reactor Vessel Level indicates 29%

x All Reactor Coolant Pumps have been tripped x Representative CET temperature is 507°F x Hot leg temperatures indicate 506°F x Cold leg temperatures indicate 504°F x Steam Generator Pressures are 900 psia x Narrow Range Steam Generator Levels are 38% and rising x Total HPSI Flow indicates 360 gpm x Total LPSI flow indicates 0 gpm What is the status of natural circulation?

A. Single phase natural circulation is occurring.

B. Two phase natural circulation is occurring.

C. Reflux boiling is occurring.

D. Natural circulation has stopped.

Question # 2 Revision: 0 KA #: 000009 EK2.03 Tier 1 Group 1 Small Break LOCA Knowledge of the interrelations between the small break LOCA and the following: S/Gs Importance: 3 / 3.3 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

STATE from memory the four indications used to verify the development of Subcooled Natural Circulation.

EXPLANATION:

Choice D is correct because primary pressure is below the S/G pressures and there is no heat sink to support natural circulation. The distractors are the three modes of natural circulation.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: 000009 EK2.03 Bank Ref #: N/A LP# / Objective: 0718-13 03.05 Exam Level: RO-5 Cognitive Level: HIGH Source: NEW

Reference:

LP 07-15-23 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 003 The following plant conditions exist:

x The plant is heating up following a refueling outage x The RCS cold leg temperatures indicate 505°F x Three Reactor Coolant Pumps are operating x The fourth RCP is about to be started In these conditions, a precaution in OI-RC-9, "Reactor Coolant Pump Operation,"

states "WHEN a RCP Motor is cold, THEN do NOT attempt more than two starts in succession."

What is the reason for this precaution?

A. To prevent damage to the RCP seals.

B. To prevent damage to the RCP bearings.

C. To prevent damage to the RCP motor windings.

D. To prevent damage to the RCP oil lift pumps.

Question # 3 Revision: 0 KA #: 000015 AK3.01 Tier 1 Group 1 Reactor Coolant Pump Malfunctions Knowledge of the reasons for the following responses as they apply to the Reactor Coolant Pump Malfunctions (Loss of RC Flow) : Potential damage from high winding and/or bearing temperatures Importance: 2.5 / 3.1 CFR Number: 5.41(b)(3)

Mechanical components and design features of the reactor primary system.

Fort Calhoun Objective:

DISCUSS the prerequisites and precautions for operating an RCP.

EXPLANATION:

Choice C is correct per the procedure. The distractors are all items that can be affected by repetitive pump starts.

KA#: 000015 AK3.01 Bank Ref #: N/A LP# / Objective: 0711-20 03.02A Exam Level: RO-3 Cognitive Level: LOW Source: NEW

Reference:

OI-RC-9 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 004 The first step of AOP-33, "CVCS LEAK," directs that all Charging Pump Control Switches be placed in PULL-TO-LOCK and that the following valves be closed:

x TCV-202, Letdown Isolation Valve x HCV-204, Letdown Isolation Valve x HCV-238, Loop 1 Charging Isolation x HCV-239, Loop 2 Charging Isolation x HCV-240, PZR Auxiliary Spray Isolation Valve x HCV-249, PZR Auxiliary Spray Isolation Valve Why is Charging Line Stop Valve, HCV-247, not closed during this step?

A. Because it is normally closed.

B. Because it is interlocked to automatically close when HCV-204 is closed.

C. To maintain a Charging Header Relief Path to the Volume Control Tank.

D. To maintain a Charging Header Relief Path to the RCS Loops.

Question # 4 Revision: 0 KA #: 000022 2.4.06 Tier 1 Group 1 Loss of Reactor Coolant Makeup Knowledge EOP mitigation strategies.

Importance: 3.7 / 4.7 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Given the caution statements and/or notes listed in this AOP, EXPLAIN the reason for EXPLANATION:

D is the correct answer per the reference. The distractors are all credible, but incorrect reasons.

KA#: 000022 2.4.06 Bank Ref #: N/A LP# / Objective: 0717-33 01.05 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

TBD-AOP-33 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 005 Given the following plant conditions:

x The plant is on shutdown cooling x No Reactor Coolant Pumps are operating x Pressurizer Manways are installed How will a loss of power to RCS Pressure Controller, PC-118 affect Loop Suction Valves, HCV-347 and HCV-348?

A. HCV-347 and HCV-348 will both close.

B. Only HCV-347 will close.

C. Only HCV-348 will close.

D. HCV-347 and HCV-348 will remain open unless power is also lost to PC-115.

Question # 5 Revision: 0 KA #: 000025 AA1.10 Tier 1 Group 1 Loss of Residual Heat Removal System Ability to operate and / or monitor the following as they apply to the Loss of Residual Heat Removal System: LPI pump suction valve and discharge valve indicators Importance: 3.1 / 2.9 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Given a current copy of OI-SC-1, explain the major steps, prerequisites and precautions for placing the Shutdown Cooling System in service.

EXPLANATION:

A loss of power to PC-118 will close both HCV-347 and HCV-348. "A" is correct and the distractors are incorrect.

KA#: 000025 AA1.10 Bank Ref #: 07-11-22 025 LP# / Objective: 0711-22 01.04 Exam Level: RO-7 Cognitive Level: LOW Source: BANK

Reference:

STM 15 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 006 Given the following plant conditions:

x The plant is operating at full power x Volume Control tank level is lowering x CCW Surge Tank level and pressure are rising x Rising count rate indicated by CCW radiation monitor, RM-053 Which one of the following events would cause these indications?

A. A tube leak in the CVCS regenerative heat exchanger.

B. A leak in a reactor coolant pump seal cooler.

C. A tube leak in the spent fuel pool heat exchanger.

D. A leak in a Raw Water/CCW heat exchanger.

Question # 6 Revision: 0 KA #: 000026 AA1.05 Tier 1 Group 1 Loss of Component Cooling Water Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water:

The CCWS surge tank, including level control and level alarms, and radiation alarm Importance: 3.1 / 3.1 CFR Number: 5.41(b)(11)

Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Fort Calhoun Objective:

EXPLAIN conditions that indicate leakage in or out of the CCW System.

EXPLANATION:

B would result in a leak of RCS liquid to CCW producing the conditions in the stem. Choice A is incorrect because there is no CCW interface (CVCS letdown heat exchanger would be correct), C & D are incorrect because the leakage would be from the CCW system.

KA#: 000026 AA1.05 Bank Ref #: 07-11-06 LP# / Objective: 0711-06 06.01 Exam Level: RO-11 Cognitive Level: HIGH Source: NRC 1997 REWORD

Reference:

AOP-22 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 007 The reactor tripped 20 minutes ago. The following conditions are observed:

x "PRESSURIZER PRESSURE OFF NORMAL HI-LO" channel X and Y are in alarm x PRC-103Y (controlling channel) indicates 2175 psia and slowly rising x PRC-103X indicates 1980 psia and slowly lowering x All backup heaters are in auto and deenergized x LRC-101Y (controlling channel) indicates 48% and steady.

x LRC-101X indicates 47% and steady x Letdown flow is 36 gpm x One charging pump is running x Tcold indicates 533°F, Thot indicates 534°F, both are stable What action should be taken to restore RCS pressure to normal?

A. Select PRC-103X as the controlling pressure channel.

B. Take manual control of PRC-103Y to lower pressurizer pressure.

C. Select LRC-101X as the controlling level channel.

D. Place pressurizer heater cutout switch, HC-101-1, in channel "X."

Question # 7 Revision: 0 KA #: 000027 AK2.03 Tier 1 Group 1 Pressurizer Pressure Control System Malfunction Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners Importance: 2.6 / 2.8 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

When given specific plant conditions, EXPLAIN operating principles to predict response of Reactor Coolant System (RCS) Instrumentation.

EXPLANATION:

A is correct. These conditions would result if PRC-103Y failed high. B is incorrect because PRC-103Y has failed. C&D are incorrect because there is no problem with level. If either level channel had failed low, all backup heaters would deenergize. Choice D, would have no effect CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: 000027 AK2.03 Bank Ref #: 07-11-20 156 LP# / Objective: 0711-20 04.00 Exam Level: RO-7 Cognitive Level: HIGH Source: MODIFIED

Reference:

STM 37 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 008 Depressing the manual reactor trip pushbutton on CB-4 de-energizes the CEDM clutch power supplies by ________________.

A. De-energizing the "M" coils which opens the "M" contacts.

B. Energizing the "M" coils which opens the "M" contacts.

C. De-energizing the undervoltage coils for the CEDM clutch power supply circuit breakers causing them to open.

D. Energizing the shunt coils for the CEDM clutch power supply circuit breakers causing them to open.

Question # 8 Revision: 0 KA #: 000029 2.1.31 Tier 1 Group 1 Anticipated Transient Without Scram (ATWS)

Ability to locate control room switches, controls and indications and to determine that they correctly reflecting the desired plant lineup.

Importance: 4.6 / 4.3 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Given a simplified diagram of the RPS trip paths, EXPLAIN how the "M" coil contacts are: Opened to initiate a reactor trip EXPLANATION:

Pushing the Reactor trip pushbutton on CB-4, de-energizes the "M" coils, opening "M" contacts which removes power to the clutch power supplies. "A" is correct. "B" is incorrect because energizing the coils closes the contacts. "C" and "D" are wrong because the CB-4 pushbutton has no affect on the clutch power supply circuit breakers, (although "Diverse Scram does open them)

KA#: 000029 2.1.31 Bank Ref #: 07-12-25 009 LP# / Objective: 0712-25 01.15 Exam Level: RO-7 Cognitive Level: HIGH Source: BANK

Reference:

STM 38 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 009 What automatic action is caused by a high radiation alarm on Condenser Off-gas Radiation Monitor, RM-057?

A. Blowdown isolation valves, HCV-1387A/B and HCV-1388A/B, close to terminate a radiation release to the environment.

B. Condenser off-gas is routed through the hydrogen purge filters and the auxiliary building stack to reduce the Iodine and Particulate release to the environment.

C. 6th stage extraction feed to Auxiliary Steam Valve, RCV-978, closes to prevent contamination of the Auxiliary Steam System.

D. Sample Valves from the Main Steam Line to Main Steam Line Radiation Monitor, RM-064, open to provide additional monitoring of a release to the environment.

Question # 9 Revision: 0 KA #: 000038 EK3.04 Tier 1 Group 1 Steam Generator Tube Rupture Knowledge of the reasons for the following responses as the apply to the SGTR: Automatic actions provided by each PRM Importance: 3.9 / 4.1 CFR Number: 5.41(b)(11)

Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Fort Calhoun Objective:

LIST radiation monitors with automatic actuations and STATE the automatic actuations that occur.

EXPLANATION:

RM-057 High Radiation isolates RCV-978, thus "C" is the correct answer. "A" is incorrect, because high radiation on RM-054A/B isolates blowdown. Choices "B" and "D" are both manual actions. They are valid distractors because they are manual actions taken in response to a steam generator tube rupture.

KA#: 000038 EK3.04 Bank Ref #: 07-12-03 010 LP# / Objective: 0712-03 04.01 Exam Level: RO-11 Cognitive Level: LOW Source: NRC 1999 EXAM

Reference:

STM 33 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 010 The plant was operating at full power when the following sequence of events occurred:

x Level and pressure in Steam Generator, RC-2A, began lowering.

x Indicated Feedwater flow to RC-2A increased.

x Indicated Feedwater flow to RC-2B decreased.

x The reactor tripped followed by PPLS, CPHS,SGLS, and SGIS.

x All Safeguards Equipment operated as designed.

x Pressure and level in RC-2A continued to lower after both MSIVs closed.

Which one of the following break locations would cause these indications?

A. A break in the "A" feedwater line downstream of the FW Check Valve.

B. A break in the "A" feedwater line between the containment penetration and the FW Check Valve.

C. A break in the "B" Auxiliary feedwater line downstream of the FW Check Valve.

D. A break in the "A" Main steam line in room 81, upstream of the MSIV.

Question # 10 Revision: 0 KA #: 000054 AK1.01 Tier 1 Group 1 Loss of Main Feedwater Knowledge of the operational implications of the following concepts as they apply to Loss of Main Feedwater (MFW): MFW line break depressurizes the S/G (similar to a steam line break)

Importance: 4.1 / 4.3 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

EXPLAIN the plant response to an Excessive Heat Removal Event.

EXPLANATION:

The stem conditions are indications of a feedwater line break between the FW check valve and the "A" S/G, the S/G blows down much like a steam line break. Thus choice "A" is correct. Choice "B" is incorrect, because RC-2A pressure would not lower. "C" is incorrect because S/G "A" would not depressurize.

Choice "D" is incorrect because CPHS would not have actuated for a steam line break in room 81.

KA#: 000054 AK1.01 Bank Ref #: N/A LP# / Objective: 0715-20 01.00 Exam Level: RO-5 Cognitive Level: HIGH Source: NEW

Reference:

LP 07-15-12 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 011 The following plant conditions exist:

x The reactor and turbine have tripped x 4160 V buses 1A1, 1A2, 1A3, and 1A4 indicate 0 Volts x Diesel Generator D-1 is out of service x Diesel Generator D-2 failed to start automatically x 13.8 KV is available to the site x 345 KV is available to the site Which one of the following EOP/AOP Attachments or Floating Steps contains actions that must be taken within 15 minutes per EOP-00, "STANDARD POST TRIP ACTIONS?"

A. EOP/AOP Floating Step Y, "345 KV Backfeed."

B. EOP/AOP Attachment 5, "Energizing 480 V Buses From 13.8 KV."

C. EOP/AOP Attachment 6, "Minimizing DC Loads."

D. EOP/AOP Attachment 13, "Emergency Start of Diesel Generator D-2."

Question # 11 Revision: 0 KA #: 000055 2.4.01 Tier 1 Group 1 Station Blackout Knowledge of EOP entry conditions and immediate action steps.

Importance: 4.6 / 4.8 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

GIVEN a copy of Attachment 6, EXPLAIN the steps necessary to minimize DC loads.

EXPLANATION:

Step 1 of the Minimize DC loads floating step must be taken within 15 minutes of a station blackout to ensure the batteries will have the capability to power plant instrumentation for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. Therefore, "C" is the correct answer. The distractors are all actions that may be taken to mitigate a station blackout and are, therefore, credible. This is an RO question because directing DC loads to be minimized is a contingency action in EOP-00.

KA#: 000055 2.4.01 Bank Ref #: N/A LP# / Objective: 0718-17 02.03 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

EOP-07 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 012 Given the following plant conditions:

x The reactor and turbine tripped 20 minutes ago x 4160 volt buses 1A1 and 1A2 indicate 0 volts x 4160 volt buses 1A3 and 1A4 indicate 4150 volts x Diesel Generators D-1 and D-2 are running and loaded.

x Condenser Vacuum indicates 0.5 inches Hg x Both MSIVs are open x Atmospheric Relief Valve, HCV-1040, is isolated x FW-54 is providing flow to both Steam Generators x The Operators have not opened MS-291 or MS-292 Under these conditions, what would be the expected pressure indicated on CB-10/11 by PI-941A, Main Steam Pressure?

A. 800 - 850 psia.

B. 875 - 925 psia.

C. 975 - 1025 psia.

D. 1050 - 1100 psia Question # 12 Revision: 0 KA #: 000056 AA2.86 Tier 1 Group 1 Loss of Off-Site Power Ability to determine and interpret the following as they apply to the Loss of Offsite Power: Main steam pressure meter scale Importance: 2.7 / 2.7 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

EXPLAIN when the atmospheric dump would be used instead of the steam dump and bypass valves.

EXPLANATION:

Following a loss of offsite power with no operator action, S/G pressure will be controlled at approximately 1000 psia due to automatic cycling of MS-291/292 making "C" the correct answer. "B" would be the correct answer if condenser vacuum were available. "D" is the set pressure of some other safety valves, but they would not be cycling 20 minutes after the trip. "A" is the pressure range for normal full power operation.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: 000056 AA2.86 Bank Ref #: N/A LP# / Objective: 0711-17 04.01 Exam Level: RO-5 Cognitive Level: HIGH Source: NEW

Reference:

STM 25 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 013 Given the following plant conditions:

x The plant is operating at Full Power x The "INVERTER A TROUBLE" and "INSTRUMENT BUS A LOW VOLTAGE/GROUND" annunciators are in alarm x Instrument Bus A voltage on AI-40A indicates 0 volts AOP-16,Section II, "Loss of Instrument Bus AI-40A," directs that HC-111/121, REACTOR REG SYSTEM CHANNAL SELECTOR SWITCH, be placed in the "B" position.

What control system will be affected if this action is not taken?

A. Condenser Steam Dump Valve Control.

B. Steam Generator Level Control.

C. Pressurizer Level Control.

D. Pressurizer Pressure Control.

Question # 13 Revision: 0 KA #: 000057 AK3.01 Tier 1 Group 1 Loss of Vital AC Electrical Instrument Bus Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: Actions contained in EOP for loss of vital ac electrical instrument bus Importance: 4.1 / 4.4 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Describe how the plant responds to a loss of instrument bus power in terms of how specific equipment is affected and how it affects overall plant operation and reliability.

EXPLANATION:

The Reactor Regulating System provides input to both the Pressurizer Level Control System and The Steam Dump and Bypass Valve Control System, however the HC-111/121 switch only affects Steam Dump and Bypass Control. Therefore, "A" is the correct answer and "C" is a valid distractor. "B" and "D" are both Control Systems with many inputs and are also valid distractors.

KA#: 000057 AK3.01 Bank Ref #: N/A LP# / Objective: 0717-16 01.02 Exam Level: RO-7 Cognitive Level: HIGH Source: NEW

Reference:

STM 36 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 014 The following conditions exist:

x A loss of both DC buses has resulted in a loss of plant instrumentation, 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br /> after a station blackout.

x The DC buses were lost because the batteries fully discharged.

x 13.8 KV and 161KV have just been restored to the switchyard.

What actions would result in restoring normal DC control power required to enable energizing bus 1A3?

A. Use 13.8 KV to supply battery charger # 1. Use battery charger #1 to power DC bus #1 B. Use 13.8 KV to supply battery charger # 2. Use battery charger #2 to power DC bus #2 C. Use 13.8 KV to supply battery charger # 3. Use battery charger #3 to power DC bus #1 D. Use 13.8 KV to supply battery charger # 3. Use battery charger #3 to power DC bus #2.

Question # 14 Revision: 0 KA #: 000058 AK1.01 Tier 1 Group 1 Loss of DC Power Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power:

Battery charger equipment and instrumentation Importance: 2.8 / 3.1 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Explain the principles of emergency operation of the 480 VAC Electrical Distribution System in terms of major parameters, alarms and control devices.

EXPLANATION:

Normal DC Control Power for bus 1A3 comes from DC bus#1. 13.8 KV can only be used to power battery charger #3, Thus "C" is correct and the other choices are credible, but incorrect.

KA#: 000058 AK1.01 Bank Ref #: 07-13-03 013 LP# / Objective: 0713-03 01.07 Exam Level: RO-10 Cognitive Level: HIGH Source: MODIFIED

Reference:

STM 14 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 015 Given the following conditions:

x The Plant is operating at Full Power x River Level is 990 feet x Raw Water Pumps AC-10A and AC-10B are operating x The "RAW WATER SUPPLY HEADER FLOW LOW" annunciator is in alarm x The "RAW WATER STRAINER DIFFERENTIAL HI " annunciators are in alarm x The "SCREEN OR GRID A AND B DIFFERENTIAL HI" annunciator is in alarm x The red pressure indicating lights for AC-10A and AC-10B are lit x The 10 psig and 25 psig pressure indicating lights on the crosstie upstream of the flow transmitters are lit x Downstream of the flow transmitters, the 10 psig lights are lit but the 25 psig lights are off.

Which one of the following is a possible cause of the Low Flow alarm?

A. Clogged CW traveling screens.

B. Clogged RW discharge strainers.

C. A leak in the RW backup header in room 18.

D. Erosion of the RW pump impellers.

Question # 15 Revision: 0 KA #: 000062 AA2.02 Tier 1 Group 1 Loss of Nuclear Service Water Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The cause of possible SWS loss Importance: 2.9 / 3.6 CFR Number: 5.41(b)(4)

Secondary coolant and auxiliary systems that affect the facility.

Fort Calhoun Objective:

Use the Loss of Raw Water Procedure to mitigate the consequences of a loss of cooling to Component Coolin Water System or a leak in the Raw Water System.

EXPLANATION:

The crosstie upstream of the flow transmitters is also upstream of the strainers, the indicating lights downstream of the flow transmitters are also downstream of the strainers. Choice "B" is correct because it shows a high DP accross the strainers. The distractors are all valid causes of low flow and/or pressure, but would not provide all of the indications in the stem.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: 000062 AA2.02 Bank Ref #: N/A LP# / Objective: 0717-18 01.00 Exam Level: RO-4 Cognitive Level: HIGH Source: NEW

Reference:

STM 35 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 016 Which of the following is designed to protect safeguards equipment motors and their associated 480 V buses against starting with a degraded voltage condition on the system grid?

A. OPLS sensor circuit B. Undervoltage Load Shed C. 183/MES switch on D-2 panel D. Fast transfer circuitry Question # 16 Revision: 0 KA #: 000077 AK2.01 Tier 1 Group 1 Generator Voltage and Electric Grid Disturbances Knowledge of the interrelations between Generator Voltage and Electric Grid Disturbances and the following: Motors Importance: 3.1 / 3.2 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Explain how the system responds automatically to malfunctions.

EXPLANATION:

OPLS is designed for this purpose, choice "A" is correct. The distractors all affect safeguards eqipment during various electrical failures or conditions.

KA#: 000077 AK2.01 Bank Ref #: 07-13-02 021 LP# / Objective: 0713-02 01.09 Exam Level: RO-7 Cognitive Level: LOW Source: BANK

Reference:

STM 19 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 017 A reactor trip has just occurred and the Standard Post Trip actions are being performed. The following annunciators are in alarm:

x "ROD POSITION DEVIATION LOW LIMIT" x "ROD POSITION DEVIATION LOW-LOW LIMIT" x "PLANT AIR PRESS LO" x "FEEDWATER CONTROL STEAM GENERATOR RC-2A LEVEL LO-LO" x "FEEDWATER CONTROL STEAM GENERATOR RC-2B LEVEL LO-LO" x "PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL X" x "PRESSURIZER PRESSURE OFF NORMAL HI-LO CHANNEL Y" The following indications are noted:

x All trippable CEAs are fully inserted except for B-15 which is fully withdrawn x Instrument Air Pressure is 85 psig and stable x WR Levels in both Steam Generators indicate 85% and stable x Pressurizer Pressure indicates 2060 psia and stable Which one of the following contingency actions should be taken per EOP-00, "STANDARD POST TRIP ACTIONS?

A. Emergency Boration B. Start an additional Air Compressor C. Start an AFW pump.

D. Place all Pressurizer Backup Heaters in the "ON" position.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 17 Revision: 0 KA #: CE-E02 EK1.03 Tier 1 Group 1 Reactor Trip Recovery Knowledge of the operational implications of the following concepts as they apply to the (Reactor Trip Recovery) Annunciators and conditions indicating signals, and remedial actions associated with the (Reactor Trip Recovery).

Importance: 3 / 3.4 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

GIVEN a set of plant conditions and a copy of the EOP resource Assessment Trees, DETERMINE the correct success path for any of the following safety functions:

EXPLANATION:

Instrument air pressure is less than 90 psig, therefore the EOP-00 contingency action of starting an air compressor should be taken. The other parameters are within their acceptance criteria.

KA#: CE-E02 EK1.03 Bank Ref #: 07-18-10 057 LP# / Objective: 0718-10 01.05 Exam Level: RO-10 Cognitive Level: LOW Source: NRC 02 EXAM

Reference:

EOP-00 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 018 Given the following initial plant conditions:

x A plant heatup is in progress x Steam generator pressure is 610 psia x Pressurizer Pressure is 1710 psia x No Operator action was taken to unblock PPLS or SGLS A steam line break in Room 81 then causes steam flow to rise, steam generator pressure to drop to 550 psia and pressurizer pressure to drop to 1570 psia Which of the following ESF actuations would be expected?

A. CPHS B. PPLS C. SGLS D. CSAS Question # 18 Revision: 0 KA #: CE-E05 EA1.01 Tier 1 Group 1 Excess Steam Demand Ability to operate and / or monitor the following as they apply to the (Excess Steam Demand)

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Importance: 3.9 / 4.2 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

DESCRIBE the operation of the Engineered Safeguards Control System during normal and emergency conditions.

EXPLANATION:

PPLS will have automatically unblocked above 1700 psia pressurizer pressure, thus when RCS pressure drops to 1600 psia, PPLS will actuate, choice "B" is correct. CPHS and CSAS will not actuate because the break is outside of containment. SGLS will not actuate unless S/G pressure falls to 500 psia. Thus, the distractors are all incorrect.

KA#: CE-E05 EA1.01 Bank Ref #: 07-12-14 093 LP# / Objective: 0712-14 02.00 Exam Level: RO-7 Cognitive Level: HIGH Source: MODIFIED

Reference:

STM 19 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 019 Given the following plant conditions:

x The plant is operating at 60% power x "ROD DROP NUCLEAR INSTRUMENTATION CHANNEL" is in alarm x "ROD POSITION DEVIATION LOW LIMIT" is in alarm x "ROD POSITION DEVIATION LOW-LOW LIMIT" is in alarm x "PDIL GR 4 COMPUTER" is in alarm x The Group selector switch is in "Group 4" position x The Group 4 synchro indicates 118 inches x SCEAPIS indicates that CEA # 1 is at 18 inches x The white CEA mimic light for CEA 1 is lit Which one of the following would produce these indications?

A. The Primary CEA position indication system (synchro) has failed B. The Secondary CEA position indication system (DCS SCEAPIS) has failed C. A CEA has fully inserted into the core.

D. A CEA has partially inserted into the core Question # 19 Revision: 0 KA #: 000003 AA2.01 Tier 1 Group 2 Dropped Control Rod Ability to determine and interpret the following as they apply to the Dropped Control Rod: Rod position indication to actual rod position Importance: 3.7 / 3.9 CFR Number: 5.41(b)(6)

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

Fort Calhoun Objective:

Use the CEA and Control System Malfunctions Procedure to mitigate the consequences of a malfunction of a CEA, the CEA control system or CEA position indication.

EXPLANATION:

These are indications of a partial rod drop of CEA #1 to 18 inches. Choice "D" is correct. Choices "A" and "B" are incorrect because the "Rod Drop Nuclear Instrumentaion" alarm comes in when reactor power drops 8% in 8 seconds, therefore a CEA did insert. Choice "C" is incorrect because the white mimic light would be off for a fully inserted CEA.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: 000003 AA2.01 Bank Ref #: N/A LP# / Objective: 0717-02 01.00 Exam Level: RO-6 Cognitive Level: HIGH Source: NEW

Reference:

STMS 11&29 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 020 The following plant conditions exist:

x Spent fuel movement is in progress inside containment x Control Room HVAC unit, VA-46A, has tripped and cannot be restarted Which one of the following actions will be required by Technical Specifications in order to continue fuel movement?

A. Place Control Room HVAC unit, VA-46B, in operation in the "Filtered Air Mode" immediately.

B. Place Control Room HVAC unit, VA-46B, in operation in the "Filtered Air Mode" within one hour.

C. Place Control Room HVAC unit, VA-46B, in operation in the "Recirculation Mode" immediately.

D. Place Control Room HVAC unit, VA-46B, in operation in the "Recirculation Mode" within one hour.

Question # 20 Revision: 0 KA #: 000036 2.2.39 Tier 1 Group 2 Fuel Handling Incidents Knowledge of less than or equal to one hour Technical Specification action statements for systems.

Importance: 3.9 / 4.5 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

State the Technical Specifications, and the bases associated with the Contol Room Ventilation System.

EXPLANATION:

Control Room HVAC must be in Filtered Air Mode whenever fuel is being moved. There is no one hour limit. Therefore choice "A" is correct. Choice "B" is incorrect because there is no one hour limit. Choices "C" and "D" are incorrect because Filtered Air mode is required.

KA#: 000036 2.2.39 Bank Ref #: N/A LP# / Objective: 0714-06 01.09 Exam Level: RO-5 Cognitive Level: LOW Source: NEW

Reference:

TS 2.8.2(4) Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 021 The following measurements of primary to secondary leakage were made:

Day RC-2A Leakage RC-2B Leakage Monday 0.00 gpm 0.05 gpm Tuesday 0.05 gpm 0.10 gpm Wednesday 0.10 gpm 0.15 gpm Thursday 0.15 gpm 0.20 gpm When was entry into Technical Specification 2.1.4 first required due to primary to secondary leakage?

A. Monday B. Tuesday C. Wednesday D. Thursday Question # 21 Revision: 0 KA #: 000037 2.2.22 Tier 1 Group 2 Steam Generator Tube Leak Knowledge of limiting conditions for operations and safety limits.

Importance: 4 / 4.7 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

Describe the Technical Specification LCO that is challenged by a leak in the Reactor Coolant System.

EXPLANATION:

The Tech Spec Limit is 150 gpd (0.104 gpm) through any one S/G.Therefore the limit was first exceeded on Wednesday when the leakage in RC-2B exceeded 150 gpd.

KA#: 000037 2.2.22 Bank Ref #: N/A LP# / Objective: 0717-22 01.06 Exam Level: RO-5 Cognitive Level: HIGH Source: NEW

Reference:

TS 2.1.4 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 022 Given that the following plant conditions:

x "AUX BLDG SUMP AREA 23 HI or LOW LEVEL" alarm x "SPENT REGEN HOLDUP TANK WD-13A HI OR LOW LEVEL" alarm x "RM-062 AUX BLDG VENT STACK HIGH RADIATION" alarm x VIAS has actuated x Aux Building Operator reports high level in room 23 sump and low level in WD-13A x AOP-09, "HIGH RADIOACTIVITY" has been entered and actions taken x All auxiliary building exhaust fans have tripped and cannot be restarted Why does AOP-09 direct that Security unlock doors between the Auxiliary building and Radwaste Building and that the doors be opened?

A. To allow personnel to exit the Auxiliary Building faster B. To allow hoses to be run to transfer water from room 23's sump C. To ensure adequate heat removal from the Auxiliary Building D. To ensure that the release is monitored by RM-043 Question # 22 Revision: 0 KA #: 000059 AK3.04 Tier 1 Group 2 Accidental Liquid Radwaste Release Knowledge of the reasons for the following responses as they apply to the Accidental Liquid Radwaste Release: Actions contained in EOP for accidental liquid radioactive-waste release Importance: 3.8 / 4.3 CFR Number: 5.41(b)(11)

Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Fort Calhoun Objective:

Use the High Radioactivity Procedure to mitigate the consequences of unplanned or uncontrolled high radiation levels in any area of the plant.

EXPLANATION:

TBD-AOP-09 states that this action is taken to ensure that the release is monitored by RM-043, choice "D" is the correct answer. The distractors are credible alternatives for opening the doors.

KA#: 000059 AK3.04 Bank Ref #: N/A LP# / Objective: 0717-09 01.00 Exam Level: RO-11 Cognitive Level: LOW Source: NEW

Reference:

TBD-AOP-09 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 023 A Radiation Area Monitor is showing the following indications in the Control Room:

x The RANGE alarm red light is lit x The panel display indicates "EE.EEE mR/hr" What would cause these indication?

A. The dose rate at the detector location is less than 0.1 mR/hr.

B. The dose rate at the detector is greater than 1 x 107 mR/hr.

C. The detectors high voltage power supply has failed.

D. The connection between the local ratemeter and the control room ratemeter has failed Question # 23 Revision: 0 KA #: 000061 AK1.01 Tier 1 Group 2 Area Radiation Monitoring (ARM) System Alarms Knowledge of the operational implications of the following concepts as they apply to Area Radiation Monitoring (ARM) System Alarms: Detector limitations Importance: 2.5 / 2.9 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN the characteristics of the components which make up the Radiation Monitoring System.

EXPLANATION:

This is an indication that the dose rate exceeds the range of the detector, "B" is the correct answer. "A" is incorrect because the display will indicate 0.0 in this situation. "C" and "D" will result in a trouble alarm.

KA#: 000061 AK1.01 Bank Ref #: N/A LP# / Objective: 0712-03 02.00 Exam Level: RO-7 Cognitive Level: LOW Source: NEW

Reference:

STM 33 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 024 The following plant conditions exist:

x The reactor has been tripped and Control Room evacuated due to a fire.

x Auxiliary Feedwater control is being established at AI-179.

x Both transfer switches (43/RC-2A & 43/RC-2B) was taken to local but relay 43X/RC-2A failed to trip.

How will this failure affect the ability to feed the Steam Generators from AI-179?

A. The Operators will not be able to feed either Steam Generator.

B. The Operators will be unable to feed Steam Generator RC-2A.

C. The Operator will be unable to feed Steam Generator RC-2B D. The Operators will be able to feed both Steam Generators.

Question # 24 Revision: 0 KA #: 000068 AK2.03 Tier 1 Group 2 Control Room Evacuation Knowledge of the interrelations between the Control Room Evacuation and the following: Controllers and positioners Importance: 2.9 / 3.1 CFR Number: 5.41(b)(7)

Design, components, and function of control and safety systems, including instrumentation, signals, interlocks, failure mods, and automatic and manual features Fort Calhoun Objective: EXPLAIN the operation of the auxiliary relays (43X/RC2A and 43X/RC-2B) and transfer switches (43/RC-2A and 43/RC-2B) on AI-179.

EXPLANATION: There is no ability to control FCV-1107A, AFW flow control to RC-2A. RC-2B can still be fed KA#: 000076 AK2.01 Bank Ref #: 0712-01 024 LP# / Objective: 0712-01 01.05 Exam Level: RO-11 Cognitive Level: LOW Source: BANK

Reference:

AOP-06 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 025 Given the following plant conditions:

x The plant was operating at full power x A steam line break occurred inside containment x Offsite power was lost coincident with the steam line break x Both Diesel Generators started and loaded as designed Why does EOP-05, "UNCONTROLLED HEAT EXTRACTION," direct that steam flow be initiated from the unaffected steam generator prior to dryout of the affected steam generator?

A. To ensure continuous heat removal for the establishment of natural circulation.

B. To promote reverse heat transfer from the affected steam generator which results in a lower containment pressure.

C. To minimize repressurization of the reactor coolant system which could result in pressurized thermal shock.

D. To minimize the difference between cold leg temperatures which limits the asymetric core power distribution during a potential return to criticality.

Question # 25 Revision: 0 KA #: CE-A11 AK1.03 Tier 1 Group 2 RCS Overcooling Knowledge of the operational implications of the following concepts as they apply to the (RCS Overcooling) Annunciators and conditions indicating signals, and remedial actions associated with the (RCS Overcooling).

Importance: 3 / 3.2 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

EXPLAIN the major startegy used to mitigate the consequences of an UHE.

EXPLANATION:

"C" is the correct answer according to TBD-EOP-05. The distractors are plausible explanations for establishing heat removal. "A", a heat sink is needed for natural circulation and the stem states that a loss of offsite power has occurred. "B" The stem states that the steamline break is inside containment. "D" reactivity addition is a concern during a steam line break and Fort Calhoun has an asymetric S/G trip.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: CE-A11 AK1.03 Bank Ref #: N/A LP# / Objective: 0718-15 01.01 Exam Level: RO-5 Cognitive Level: HIGH Source: NEW

Reference:

TBD-EOP-05 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 026 Given the following plant conditions:

x The plant was operating at full power x CCW flow to the RCPs was lost x The Reactor was tripped and then the RCPs were all tripped How will RCS Cold Leg Temperature respond if the steam dump and bypass valves remain in automatic control?

A. RCS T-cold will be controlled between 515° and 525°F B. RCS T-cold will be controlled between 530° and 540°F C. RCS T-cold will be controlled betwee 545° and 555°F D. RCS T-cold will be controlled between 560° and 570°F Question # 26 Revision: 0 KA #: CE-A13 AA1.02 Tier 1 Group 2 Natural Circulation Operations Ability to operate and / or monitor the following as they apply to the (Natural Circulation Operations)

Operating behavior characteristics of the facility.

Importance: 3.1 / 3.6 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

EXPLAIN the operator actions required to monitor and maintain subcooled natural circulation.

EXPLANATION:

If steam dump and bypass valves are left in automatic, T-ave will be controlled at 535°F. Since there is a 30°F DT on natural circulation, T-cold will be about 520°F. "A" is correct the others are incorrect. "B" is the value if not on natural circulation. "C" is the value if the S/G safety vales are controlling. "D" is the value for normal full power operation.

KA#: CE-A13 AA1.02 Bank Ref #: 07-15-16 029 LP# / Objective: 0715-16 02.05 Exam Level: RO-5 Cognitive Level: HIGH Source: NRC 04 EXAM

Reference:

LP 07-15-16 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 027 The ERF computer is being used to calculate the RCS leakrate using OP-ST-RC-3001,"REACTOR COOLANT SYSTEM (RCS) LEAK RATE TEST",

Attachment 1, "ERF Computer Operable - RLR."

The leakrate test will be void if _________ during the test.

A. T-cold changes by more than 1 °F B. normal makeup to the VCT is performed C. an additional charging pump is started D. the Reactor Coolant Drain Tank is pumped down Question # 27 Revision: 0 KA #: CE-A16 AK3.02 Tier 1 Group 2 Excess RCS Leakage Knowledge of the reasons for the following responses as they apply to the (Excess RCS Leakage)

Normal, abnormal and emergency operating procedures associated with (Excess RCS Leakage).

Importance: 2.8 / 3.3 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Describe how the plant responds to a Reactor Coolant Leak in terms of how specific equipment is affected and how it affects overall plant operation and reliability.

EXPLANATION:

OP-ST-RC-3001 states that the test will be void if the RCDT is pumped down, "D" is the correct answer.

The procedure also states that the test will not be affected by normal makeup to the VCT, so "B" is incorrect. The leakrate accounts for changes in RCS temperature and charging and letdown flow even though these are parameters that Operators could use to detect excessive leakage.

KA#: CE-A16 AK3.02 Bank Ref #: 07-12-14 096 LP# / Objective: 0717-22 01.02 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

OP-ST-RC-3001 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 028 The following plant conditions exist:

xThe reactor is at 100% power xRCS pressure is 2100 psia xThe RC-3A "SEAL LEAKAGE FLOW HI" annunciator is in alarm xVCT pressure is 50 psia.

xRC-3A middle seal inlet pressure is 110 psia xRC-3A upper seal inlet pressure is 80 psia What is the status of RC-3A's seals?

A. Only the upper seal has failed B. Only the lower seal has failed C. The upper and middle seals have failed D. The lower and middle seals have failed.

Question # 28 Revision: 0 KA #: 003000 K6.02 Tier 2 Group 1 Reactor Coolant Pump System Knowledge of the effect of a loss or malfunction on the following will have on the RCPS: RCP seals and seal water supply Importance: 2.7 / 3.1 CFR Number: 5.41(b)(3)

Mechanical components and design features of the reactor primary system Fort Calhoun Objective:

EXPLAIN the operation of the RCP seal package.

KA#: 003000 K6.02 Bank Ref #: 07-11-20 014 LP# / Objective: 0711-20 01.07D Exam Level: RO-10 Cognitive Level: HIGH Source: NRC 04 EXAM

Reference:

ARP CB-1,2,3/A6 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 029 Given the following plant conditions:

x The plant is in hot shutdown x I&C technicians are calibrating CCW pressure switches x 1 CCW pump is operating with normal discharge pressure x An inadvertent safeguards signal resulted in closure of HCV-438A/B/C/D What action should be taken to restore cooling water to the Reactor Coolant Pumps?

A. Immediately PULL TO OVERIDE the control switches for HCV-438A/B/C/D.

B. Wait 30 seconds and then PULL TO OVERIDE the control switches for HCV-438A/B/C/D.

C. Reset the SIAS lockout relays, then immediately place the control switches for HCV-438A/B/C/D in the OPEN position.

D. Reset the SIAS lockout relays, wait 30 seconds, then place the control switches for HCV-438A/B/C/D in the OPEN position.

Question # 29 Revision: 0 KA #: 003000 A4.08 Tier 2 Group 1 Reactor Coolant Pump System Ability to manually operate and/or monitor in the control room: RCP cooling water supplies Importance: 3.2 / 2.9 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN the operation of controls associated with the CCW System valves operated from the Control Room.

EXPLANATION:

Choice "A" is correct, the valves can be opened immediately in PULL TO OVERRIDE. Choice "B" is a plausible but incorrect distractor because there is a 30 second delay associated with automatic closure of these valves. Choices "C" and "D" are incorrect because CIAS closes these valves, not SIAS. They are plausible because SIAS does cause other safeguards actions.

KA#: 003000 A4.08 Bank Ref #: N/A LP# / Objective: 0711-06 01.02 Exam Level: RO-7 Cognitive Level: HIGH Source: NEW

Reference:

STM 8 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 030 The plant was operating at full power with Charging Pump, CH-1A in operation.

An event caused the following plant annunciators to alarm on CB-1/2/3:

x CHARGING FLOW LOW x REGENERATIVE HEAT EXCH TUBE OUTLET TEMP HI x INTERMEDIATE LETDOWN PRESS HI-LO Which one of the following events produced these alarms?

A. Pressure Transmitter, PT-210, failed low.

B. Temperature Transmitter, TE-202 failed high.

C. CH-1A discharge relief valve failed open.

D. Letdown relief valve, CH-223, failed open.

Question # 30 Revision: 0 KA #: 004000 2.4.31 Tier 2 Group 1 Chemical and Volume Control System Knowledge of annunciator alarms, indications, or response procedures.

Importance: 4.2 / 4.1 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

Given a current copy of the Annunciator Response Procedures, EXPLAIN the alarms associated with the CVCS and the required corrective actions.

EXPLANATION:

Choice "C" results in low charging flow, high letdown temperature causing letdown isolation which leads to low letdown pressure producing these 3 alarms. Choices "A", "B" and "D" would not produce the low charging flow alarm.

KA#: 004000 2.4.31 Bank Ref #: N/A LP# / Objective: 0711-02 05.02 Exam Level: RO-5 Cognitive Level: HIGH Source: NEW

Reference:

ARP CB-123 A2 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 031 At normal RCS and CVCS pressure. Placing HC-101-3, LIMITER BYPASS SWITCH, in BYPASS allows letdown flow to be increased to:

A. 116 gpm using one letdown flow control valve, LCV-101-1 or LCV-101-2.

B. 126 gpm using one letdown flow control valve, LCV-101-1 or LCV-101-2.

C. 116 gpm using both letdown flow control valve, LCV-101-1 and LCV-101-2.

D. 126 gpm using both letdown flow control valve, LCV-101-1 and LCV-101-2.

Question # 31 Revision: 0 KA #: 004000 A1.07 Tier 2 Group 1 Chemical and Volume Control System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: Maximum specified letdown flow Importance: 2.7 / 3.1 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Explain indications available in the Control Room associated with each major component.

EXPLANATION:

Each letdown flow control valve has a hard stop that limits flow to 126 gpm per valve. The pressurizer level control system normally limits flow to 116 gpm to the selected valve. HC-101-3 bypasses the 116 gpm limit to allow 126 gpm. Therefore, choice "B" is correct and the other choices are plausible, yet incorrect.

KA#: 004000 A1.07 Bank Ref #: N/A LP# / Objective: 0711-03 01.02 Exam Level: RO-7 Cognitive Level: LOW Source: NEW

Reference:

STM 12 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 032 OI-SC-5, SHUTDOWN COOLING PURIFICATION, contains a caution to maintain shutdown cooling outlet temperature, TR-346, above 54°F when shutdown cooling purification is in service.

What is the reason for this precaution?

A. To ensure that Reactor Vessel Pressure-Temperature limits are not violated.

B. To ensure that the water temperature does not go below boric acid solubility limits.

C. To ensure that assumptions made in the shutdown margin calculation remain valid.

D. To ensure that LPSI Pump motor current limits are not exceeded.

Question # 32 Revision: 0 KA #: 005000 K5.03 Tier 2 Group 1 Residual Heat Removal System Knowledge of the operational implications of the following concepts as they apply the RHRS: Reactivity effects of RHR fill water Importance: 2.9 / 3.1 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

Given a current copy of OI-SC-1, explain the major steps, prerequisites and precautions for placing the Shutdown Cooling System in service.

EXPLANATION:

The temperature limit is to ensure assumptions in the shutdown margin calculation remain valid, choice "C" is correct. "A" is a plausible distractor because there is a 64°F limit for reactor vessel boltup. "B" is a plausible distractor because boric acid solubility decreases at lower temperatures. "D" is a plausible distractor because pump motor current increases with lower temperature water.

KA#: 005000 K5.03 Bank Ref #: N/A LP# / Objective: 0711-22 01.18 Exam Level: RO-5 Cognitive Level: LOW Source: NEW

Reference:

OI-SC-5 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 033 A small break LOCA has occurred. The following plant conditions exist:

x Pressurizer pressure is 905 psia and steady x Pressurizer level is 0%

x Reactor Vessel Level is 43%

x All charging pumps are running x All HPSI pumps failed to start x Both LPSI pumps are running x No Containment Spray Pumps are running x S/G pressures are 900 psia x Narrow Range S/G levels are 27%

x AFW is being supplied to both S/G's by FW-10 x Containment pressure is 6 psig Which one of the following actions should be taken immediately?

A. Initiate Once-through-Cooling.

B. Perform LPSI stop and throttle.

C. Initiate Containment Spray Flow.

D. Increase steam flow from the steam generators.

Question # 33 Revision: 0 KA #: 006000 K6.03 Tier 2 Group 1 Emergency Core Cooling System Knowledge of the effect of a loss or malfunction on the following will have on the ECCS: Safety Injection Pumps Importance: 3.6 / 3.9 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

EXPLAIN how the decay heat removal capacity of the break affects plant response.

EXPLANATION: RCS pressure needs to be reduced to allow the Safety Injection Tanks and LPSI pumps to be able to inject. RCS pressure is staying above S/G pressure because S/G heat removal is needed.

The way to reduce RCS pressure is to lower S/G pressure by steaming making "D" the correct answer.

Distractor "A", 27% wide range (not narrow as in the stem) is the initiation condition for once through cooling. Distractor "B" RCS pressure is above the shutoff head of the LPSI pumps, but there is a recirc flow path back to the SIRWT. Distractor "C", containment pressure is above the CPHS setpoint, but it also take SGLS to initiate Containment Spray.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: 006000 K6.03 Bank Ref #: 07-15-23 030 LP# / Objective: 0715-23 01.02 Exam Level: RO-10 Cognitive Level: HIGH Source: NRC 2005 EXAM

Reference:

LP 0715-23 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 034 What method is used to form a steam bubble in the pressurizer following a refueling outage?

A. With the pressurizer level at approximately 50%, pressurizer heaters are used to heat the water to saturation. Non-condensable gases are vented to the PQT or VCT.

B. With the pressurizer level at approximately 50%, pressurizer heaters are used to heat the water to saturation. Non-condensable gases are vented to the vacuum refill system.

C. The pressurizer is filled as non-condensable gases are vented to the PQT or VCT.

With the pressurizer solid, the pressurizer heaters are used heat the water in the pressurizer. The pressurizer level is then lowered until a steam bubble forms.

D. The pressurizer is filled as non-condensable gases are vented to the vacuum refill system. With the pressurizer solid, the pressurizer heaters are used heat the water in the pressurizer. The pressurizer level is then lowered until a steam bubble forms.

Question # 34 Revision: 0 KA #: 007000 K5.02 Tier 2 Group 1 Pressurizer Relief Tank / Quench Tank System Knowledge of the operational implications of the following concepts as the apply to PRTS: Method of forming a steam bubble in the PZR Importance: 3.1 / 3.4 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

LIST the major steps for starting up the Reactor Coolant System per OP-2A.

EXPLANATION:

"A" is correct per the referenced procedures. "B" is incorrect because the vaccum refill system is used prior to establishing a steam bubble but is plausible. Choices "C" and "D" are plausible because some other plants draw a bubble starting with a solid pressurizer but not at FCS.

KA#: 007000 K5.02 Bank Ref #: 07-11-20 153 LP# / Objective: 0711-20 03.05 Exam Level: RO-10 Cognitive Level: LOW Source: NRC 2005 EXAM

Reference:

OP-2A,OI-CH-3 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 035 Given the following plant conditions:

x A loss of Raw Water has occurred and the reactor has been tripped due to high CCW temperature x EOP-00, "STANDARD POST TRIP ACTIONS," are being taken x The "LETDOWN HEAT EXCH TUBE OUTLET TEMP HI" has just alarmed What action should be taken per ARP-CB-1,2,3/A2 in response to this alarm?

A. Verify that TCV-202 has closed to prevent an inadvertent boron dilution due to boron retention in the ion exchangers.

B. Verify that TCV-202 has closed to prevent damage to the ion exchanger resin beads due to excessive temperature.

C. Verify that TCV-211-2 has switched to the BYPASS position to prevent an inadvertent boron dilution due to boron retention in the ion exchangers.

D. Verify that TCV-211-2 has switched to the BYPASS position to prevent damage to the ion exchanger resin beads due to excessive temperature.

Question # 35 Revision: 0 KA #: 008000 A2.03 Tier 2 Group 1 Component Cooling Water System Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High/low CCW temperature Importance: 3 / 3.2 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

EXPLAIN, the manual and automatic functions of control valves in the CVCS.

EXPLANATION:

Choice "D" is correct, TCV-211-2 bypasses the ion exchangers on high temperature to prevent damage to the resin beads. Choice "C" is plausible because temperature does affect boron retention by the resin beads. Choices "A" and "B" are plausible because TCV-202 is a letdown valve that will close on high temperature to protect the regenerative heat exchanger. Its setpoint is way too high (470°F) to protect resin.

KA#: 008000 A2.03 Bank Ref #: N/A LP# / Objective: 0711-02 01.02 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

ARP-CB-1,2,3/A2 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 036 Given the following plant conditions:

x The reactor tripped 20 minutes ago x "PRESSURIZER PRESSURE OFF NORMAL HI-LO" channel X and Y are in alarm x PRC-103X (controlling channel) indicates 2160 psia and stable x All backup heaters in auto and energized x LRC-101Y (controlling channel) indicates 60% and stable x LRC-101X indicates 38% and lowering slowly x LI-106 indicates 28%

x Letdown flow is 55 gpm x One charging pump is running x Tcold is 533°F, Thot is 534°F; both are stable Select the probable cause and the action to be taken?

A. Low level on LRC-101X is maintaining backup heaters on. Place the pressurizer heater cutout switch in the channel Y position B. The bistable for the backup heaters needs to be reset. Place the control switches for all B/U heaters to reset and back to auto.

C. LRC-101Y has malfunctioned causing the backup heaters to remain on. Place LRC-101X in service D. PRC-103X has malfunctioned causing the backup heaters to remain on. Place PRC-103Y in service.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 36 Revision: 0 KA #: 010000 K1.08 Tier 2 Group 1 Pressurizer Pressure Control System Knowledge of the physical connections and/or cause-effect relationships between the PZR PCS and the following systems: PZR LCS Importance: 3.2 / 3.5 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Given a current copy of ARP, EXPLAIN the alarms associated with the RCS Instrumentation System and the required actions.

EXPLANATION:

Controlling channel LRC-101Y has malfunctioned causing letdown flow to be high and lowering pressurizer level. It should be controlling to 48% at these RCS temperatures, Backup heaters are on because the level deviation is greater than 5%. "C" is correct. "A" is incorrect because low level will turn the heaters off and the setpoint is 32%. "B" is incorrect, the bistable reset is for the charging pumps. "D" is incorrect because the pressure is too high to turn on the backup heaters in AUTO.

KA#: 010000 K1.08 Bank Ref #: 07-11-20 017 LP# / Objective: 0711-20 05.04 Exam Level: RO-7 Cognitive Level: HIGH Source: NRC EXAM 2001

Reference:

STM 36 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 037 What is the basis for the RPS Asymmetric Steam Generator Transient Trip (ASGT)?

A. To ensure that peak fuel centerline temperature safety limit is not exceeded following a steam line break.

B. To ensure that peak fuel centerline temperature safety limit is not exceeded following inadvertent closure of one MSIV.

C. To ensure that DNBR stays below its limit following a steam line break.

D. To ensure that DNBR stays below its limit following inadvertent closure of one MSIV.

Question # 37 Revision: 0 KA #: 012000 K5.02 Tier 2 Group 1 Reactor Protection System Knowledge of the operational implications of the following concepts as the apply to the RPS: Power density Importance: 3.1 / 3.3 CFR Number: 5.41(b)(6)

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

Fort Calhoun Objective:

EXPLAIN the bases for each reactor trip.

EXPLANATION:

The ASGT trip is based on maintaining peak fuel centerline temperature below limits and DNBR above limits for an asymetric event such as closure on one MSIV, thus "B" is correct. "A" is incorrect, wrong event. "C" and "D" DNBR must stay above limit, not below.

KA#: 012000 K5.02 Bank Ref #: N/A LP# / Objective: 0712-25 01.04 Exam Level: RO-6 Cognitive Level: LOW Source: NEW

Reference:

STM 38 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 038 The following conditions exist:

x The plant is operating at full power x The bistables for channel "A" trip units 1,2, 9, 10 and 12 have been bypassed x The bistables for channel "D" trip units 1, 2, 9, 10 and 12 have been placed in the "tripped" condition.

Which of the following instrument failures will result in a reactor trip?

A. "A" channel NIS power range input to the RPS fails high.

B. "B" channel pressurizer pressure input to the RPS fails low.

C. "C" channel Wide range input to the RPS fails high.

D. "D" channel RCS flow input to the RPS fails low.

Question # 38 Revision: 0 KA #: 012000 A3.05 Tier 2 Group 1 Reactor Protection System Ability to monitor automatic operation of the RPS, including: Single and multiple channel trip indicators Importance: 3.6 / 3.7 CFR Number: 5.41(b)(6)

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

Fort Calhoun Objective:

State the purpose of the Power Range NI System.

EXPLANATION:

"B" will result in a TM/LP trip (2/3 logic), "A" no trip because the trip unit is bypassed (1/3 logic), "C" will not result in a trip because the high SUR trip is automatically bypassed above 15% power., "D" no trip because the trip unit is already tripped. (1/3 logic)

KA#: 012000 A3.05 Bank Ref #: 07-12-19 052 LP# / Objective: 0712-19 01.02 Exam Level: RO-6 Cognitive Level: HIGH Source: MODIFIED

Reference:

STM 38 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 039 Given the following plant conditions:

x An event has occurred that resulted in CIAS actuation x Inlet and outlet valves to Containment Cooling Coil VA-1A, HCV-400A/C have automatically closed Which one of the following could have caused these valves to close?

A. CCW pump discharge pressure switches, PCS-412 and PCS-413, failed low.

B. CCW flow transmitter from VA-1A, FT-416, failed low.

C. CCW return temperature transmitter from VA-1A, TE-420, failed high.

D. Containment cooling fan, VA-3A, tripped.

Question # 39 Revision: 0 KA #: 008000 2.1.28 Tier 2 Group 1 Component Cooling Water System Knowledge of the purpose and function of major system components and controls.

Importance: 4.1 / 4.1 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN the response of the CCW System to signals from the Engineered Safeguards Control System.

EXPLANATION:

CIAS and no CCW flow from the containment cooling coil will cause HCV-400A/C to close, choice "B". Choice "A" will cause HCV-438A/B/C/D valves to close. Choice "C" will have no affect on these valves. Tripping the containment cooling fan associated with VA-1A will also have no effect.

KA#: 008000 2.1.28 Bank Ref #: 07-11-06 011 LP# / Objective: 0711-06 01.05 Exam Level: RO-7 Cognitive Level: HIGH Source: NRC 2004 EXAM

Reference:

STM 8 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 040 Given the following plant conditions:

x The plant tripped from full power 10 minutes ago x Pressure in both steam generators is lowering due to a malfunction of the controller for turbine bypass valve, PCV-910 x No manual Operator actions are taken How will this event be mitigated?

A. Solenoid valves that enable operation of PCV-910 will deenergize when RCS Tavg falls below 535°F.

B. Solenoid valves that enable operation of PCV-910 will deenergize when Steam Generator pressure falls below 500 psia.

C. Main Steam Isolation Valves will close when RCS Tavg falls below 535°F.

D. Main Steam Isolation Valves will close when Steam Generator pressure falls below 500 psia.

Question # 40 Revision: 0 KA #: 013000 K4.03 Tier 2 Group 1 Engineered Safety Features Actuation System Knowledge of ESFAS design feature(s) and/or interlock(s) which provide for the following: Main Steam Isolation System Importance: 3.9 / 4.4 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN how each prime and backup actuation signal is developed.

EXPLANATION:

The MSIVs will close whan a SGLS occurs due to steam generator pressure being less than 500 psia.

(choice D). Solenoid valves that enable operation of TCV-909-1/2/3/4, steam dump valves, deenergize below 535°F RCS temperature. The distractors are all plausible but incorrect.

KA#: 013000 K4.03 Bank Ref #: 07-12-14 094 LP# / Objective: 0712-14 01.04 Exam Level: RO-7 Cognitive Level: LOW Source: NRC 02 EXAM

Reference:

STM 19 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 041 Given the following plant conditions:

x The plant was operating at full power x Containment cooling fans VA-3A, and VA-7D and their associated coolers were in operation x A large LOCA occurred inside containment x PPLS and CPHS actuations occurred x All CIAS lockout relays failed to actuate Assuming all other systems operate as designed, what will be the status of containment cooling 2 minutes after the LOCA?

A. VA-3A, VA-3B and VA-7D will be operating, but only VA-3A and VA-7D will have CCW flow to their associated coolers.

B. All Containment Cooling Fans will be operating, but only VA-3A and VA-7D will have CCW flow to their associated coolers.

C. All Containment Cooling Fans will be operating with CCW flow to their associated coolers.

D. All Containment Cooling Fans will be operating but there will be no CCW flow to their associated coolers.

Question # 41 Revision: 0 KA #: 022000 A3.01 Tier 2 Group 1 Containment Cooling System Ability to monitor automatic operation of the CCS, including: Initiation of safeguards mode of operation Importance: 4.1 / 4.3 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Given specific plant conditions, apply the principles of operation of the Containment Air Cooling and Filtering System to diagnose system response.

EXPLANATION:

PPLS or CPHS will start the sequencers which will send a start signal to VA-3A and VA-3B. PPLS and CPHS will cause a CSAS which will start VA-7A and VA-7B. CIAS is required to open the CCW valves to the containment coolers therefore only VA-3A and VA-7D will have cooling water flow. (choice B) "A" is incorrect because all containment cooling fans will be operating. "C" is incorrect because there will be no cooling flow to VA-3B and VA-7C. "D" is incorrect because there will be cooling flow to VA-3A and VA-7D.

All distractors are plausible because the ESF logic could have been designed that way.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: 022000 A3.01 Bank Ref #: N/A LP# / Objective: 0714-02 01.00 Exam Level: RO-7 Cognitive Level: HIGH Source: NEW

Reference:

STM 8 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 042 Given the following plant conditions:

x A Steam Line Break occurred inside containment on the steam line from S/G RC-2B x There was a coincident loss of offsite power x PPLS, CPHS and SGLS have actuated.

x Diesel Generator, D-2, failed to start Assuming that no actions are taken to cross-tie buses, what is the status of the containment spray pumps 2 minutes after CPHS actuation?

A. SI-3A is stopped, SI-3B is stopped, SI-3C stopped.

B. SI-3A is running, SI-3B is stopped, SI-3C stopped.

C. SI-3A is stopped, SI-3B is running, SI-3C stopped.

D. SI-3A is stopped, SI-3B is running, SI-3C running.

Question # 42 Revision: 0 KA #: 026000 K2.01 Tier 2 Group 1 Containment Spray System Knowledge of bus power supplies to the following: Containment spray pumps Importance: 3.4 / 3.6 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Sketch a basic single-line drawing of the ECCS, labelling all major equipment.

EXPLANATION:

SI-3A is powered from bus 1B3C which is supplied by D/G D-1. SI-3B ispowered from bus 1B4B which is supplied by failed D/G, D-2. SI-3C does not start automatically. Choice "B" is correct the other choices are incorrect.

KA#: 026000 K2.01 Bank Ref #: 07-11-22 072 LP# / Objective: 0711-22 01.01 Exam Level: RO-7 Cognitive Level: HIGH Source: MODIFIED

Reference:

STM 15 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 043 Given the following plant conditions:

x The Reactor has tripped following a loss of offsite power x Both Diesel Generators started as designed x Main Condenser Vacuum is 15" Hg and lowering x S/G Pressures are 940 psia and rising x Both MSIVs are open What is the preferred steaming path per EOP-02, "LOSS OF OFFSITE POWER /

LOSS OF FORCED CIRCULATION?"

A. Steam Dump Valves,TCV-909-1,2,3,4 B. Condenser Bypass Valve, PCV-910 C. Atmospheric Relief Valve, HCV-1040 D. Air Assisted Main Steam Safety Valves, MS-291 and MS-292 Question # 43 Revision: 0 KA #: 039000 K1.02 Tier 2 Group 1 Main and Reheat Steam System Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems: Atmospheric relief dump valves Importance: 3.3 / 3.3 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

EXPLAIN when the atmospheric dump would be used instead of the steam dump and bypass valves.

EXPLANATION:

There is insufficient condenser vacuum to operate PCV-910 or TCV-909-1.2.3.4 making "A" and "B" incorrect.

HCV-1040 is the next best choice, "C" is the corerct answer. Choice "D" is incorrect because the MSIVs are open.

KA#: 039000 K1.02 Bank Ref #: N/A LP# / Objective: 0711-17 04.01 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

EOP-02 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 044 Given the following plant conditions:

x A steamline break has occurred inside containment on the "A" S/G's steam line.

x PPLS, SGLS, SGIS, SIAS, CIAS and VIAS actuations have occurred.

What actions must be taken to open the "B" Main Steam Line Bypass Valve, HCV-1042C to establish steam flow from the "B" S/G per EOP-05, "UNCONTROLLED HEAT EXTRACTION ?"

A. Place the control switch for HCV-1042C in "OPEN."

B. Place the key operated override switch for HCV-1042C in override, then place the control switch for HCV-1042C in "OPEN."

C. Block SGLS using the key operated SGLS block switches, then place the control switch for HCV-1042C in "OPEN."

D. Block SGIS using the key operated SGIS block switches, then place the control switch for HCV-1042C in "OPEN."

Question # 44 Revision: 0 KA #: 039000 A4.01 Tier 2 Group 1 Main and Reheat Steam System Ability to manually operate and/or monitor in the control room: Main steam supply. valves Importance: 2.9 / 2.8 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN the controls and indications associated with the Main Steam System equipment manipulated from the Control Room.

EXPLANATION:

When SGLS is blocked SGIS will unblock and HCV-1042C can be opened. If there were a CPHS, then SGIS would not unblock. "C" is the correct answer. "A" is incorrect, HCV-1042C will not open with SGIS.

"B" is incorrect, there is no override switch for HCV-1042C as there is for some FW valves that close on SGIS. "D" is incorrect, there is no SGIS block switch.

KA#: 039000 A4.01 Bank Ref #: N/A LP# / Objective: 0711-17 01.02 Exam Level: RO-7 Cognitive Level: LOW Source: NEW

Reference:

EOP-05 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 045 Given the following plant conditions:

x A total loss of feedwater has occurred x Once-Through-Cooling has been established Should the condenser steam dump and bypass valves be opened to aid in primary heat removal after establishing Once-Through-Cooling? Why or why not?

A. Yes, this will lower RCS pressure to provide more HPSI flow.

B. Yes, this will ensure that natural circulation continues as long as possible.

C. No, this will result in violating RCS cooldown limits.

D. No, this will increase the probability of S/G tube thermal shock if feed flow is reestablished.

Question # 45 Revision: 0 KA #: 059000 A2.04 Tier 2 Group 1 Main Feedwater System Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Feeding a dry S/G Importance: 2.9 / 3.4 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

EXPLAIN the operator actions required during a total loss of feedwater event.

EXPLANATION:

"D" is correct per the referance. The other choices are plausible reasons to justify the answer.

KA#: 059000 A2.04 Bank Ref #: N/A LP# / Objective: 0715-17 02.03 Exam Level: RO-5 Cognitive Level: HIGH Source: NEW

Reference:

TBD-EOP-20 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 046 Given the following plant conditions:

x A Steam Generator Isolation Signal (SGIS) has isolated Feed Water to both Steam Generators.

x All Main Feedwater pumps are tripped x FW-54 will not start x Auxiliary Feedwater Pump, FW-10, is running.

Which of the following manual actions will result in water being provided to Steam Generator, RC-2As, Feed Ring?

A. Open HCV-1384, Override and Open HCV-1386 and HCV-1105 B. Open HCV-1386, Override and Open HCV-1103 and HCV-1105 C. Open HCV-1386, Override and Open HCV-1101 and HCV-1103 D. Open HCV-1384, Override and Open HCV-1385 and HCV-1105 Question # 46 Revision: 0 KA #: 061000 K1.02 Tier 2 Group 1 Auxiliary / Emergency Feedwater System Knowledge of the physical connections and/or cause-effect relationships between the AFW and the following systems: MFW System Importance: 3.4 / 3.7 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN the operation of controls located in the Control Room associated with AFW components.

EXPLANATION:

"A" is the only choice that will establish a flow path from FW-10 to S/G-A.

KA#: 061000 K1.02 Bank Ref #: 07-11-01 007 LP# / Objective: 0711-01 01.02 Exam Level: RO-7 Cognitive Level: HIGH Source: MODIFIED

Reference:

STO 04 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 047 Given the following plant conditions:

x The plant is in mode 4 with all 4160 V buses powered from 345 KV x 161 KV is unavailable due to switchyard breaker maintenance x Shutdown cooling is in operation with LPSI pump SI-1A running x Emergency Diesel Generators #1 and #2 are aligned for normal operation Assuming all systems operate as designed when 345 KV power is lost, what will be the status of the Diesel Generators?

A. Neither Diesel Generator will start.

B. Both Diesel Generators will start. D-1's output breaker will not close until SI-1A is tripped.

C. Both Diesel Generators will start. D-2's output breaker will not close until SI-1A is tripped.

D. Both Diesel Generators will start and their output breakers will close.

Question # 47 Revision:

KA #: 062000 K3.02 Tier 2 Group 1 A.C. Electrical Distribution Knowledge of the effect that a loss or malfunction of the ac distribution system will have on the following:

ED/G Importance: 4.1 / 4.4 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Explain an emergency start of the EDG. Include in your explanation the following: The conditions that will cause an auto start.

EXPLANATION:

Both D/Gs will start due to low bus voltage. The output breaker for D-1 will close even with SI-1A running."D" is correct. The distractors are plausible but incorrect.

KA#: 062000 K3.02 Bank Ref #: 07-13-05 LP# / Objective: 0713-05 01.10A Exam Level: RO-7 Cognitive Level: HIGH Source: NRC 01-2 EXAM

Reference:

STM 16 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 048 Given the following plant conditions:

x The plant was operating at full power x DC Bus 1 was lost due to a short x A Loss of Offsite Power occured x Diesel Generator D-2 started and its output breaker closed x Diesel Generator D-1 failed to start x The turbine building Operator reports all support systems for D-1 are in a normal alignment but the primary air bank pressure is 195 psig Which of the following actions will allow D-1 to be started and loaded?

A. Start the diesel from the control room using the "EMERGENCY START" button.

B. Start the diesel locally, at AI-133A.

C. Place the Air Start Motor System selector (D1-163) switch in the Number 2 position.

D. Transfer DC Control Power for DG-1 to its alternate source.

Question # 48 Revision: 0 KA #: 063000 K3.01 Tier 2 Group 1 D.C. Electrical Distribution Knowledge of the effect that a loss or malfunction of the dc electrical system will have on the following:

ED/G Importance: 3.7 / 4.1 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Explain an emergency start of the EDG. Include in your explanation the following: How emergency starts are manually inititated.

EXPLANATION:

D-1 has no DC control power, choice "D" is correct. Choices "A" and "B" will not work without DC control power. Choice "C" is incorrect, low pressure is not preventing start.

KA#: 063000 K3.01 Bank Ref #: 07-13-05 001 LP# / Objective: 0713-05 01.10H Exam Level: RO-7 Cognitive Level: HIGH Source: NRC EXAM 1995

Reference:

AOP-16 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 049 Given the following plant conditions:

x The plant was operating at full power x A loss of offsite power occurred x Both Diesel Generators started and their output breakers closed x Pumps and fans were manually placed into service per EOP-02 x 25 minutes later 161 KV power became available to the site x The CRS has directed that Bus 1A3 be powered from 161 KV per EOP/AOP Attachment 17, "Restoring Off-site Power to Bus 1A3."

x Lockout relays have been reset and Breakers 110 and 111 closed Which one of the following actions should be taken prior to closing breaker 1A33 ?

A. The breakers should be opened for all loads on the 480 volt buses powered by Bus 1A3.

B. The OPLS Test and Bypass Switches should be placed in "BYPASS."

C. Diesel Generator Breaker 1AD1 should be opened.

D. The D-1 Governor Droop Dial should be set to the "SCRIBE MARK."

Question # 49 Revision: 0 KA #: 064000 A2.14 Tier 2 Group 1 Emergency Diesel Generators Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Effects (verification) of stopping ED/G under load on isolated bus Importance: 2.7 / 2.9 CFR Number: 5.41(b)(8)

Components, capacity, and functions of emergency systems.

Fort Calhoun Objective:

GIVEN a copy of Attachment 17 or 18, EXPLAIN the steps necessary to restore off-site power to a vital 4160 V bus.

EXPLANATION:

Correct choice "D" ensures correct speed droop to protect D/G from picking up excessive load when parrelleling. Choice "A" is incorrect loads can remain on bus. "B" is incorrect, no OPLS actuation condition, "C" is incorrect, dead bus is not needed.

KA#: 064000 A2.14 Bank Ref #: N/A LP# / Objective: 0718-12 02.03 Exam Level: RO-8 Cognitive Level: LOW Source: NEW

Reference:

EOP/AOP ATTACHMENT 1 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 050 Due to rising radiation levels, an effluent pathway process radiation monitor has just reached its "ALARM" setpoint. This indicates that the offsite dose rates ________.

A. Have reached the10 CFR 20 limits for continuous release.

B. Have reached the10 CFR 20 limits for instantaneous release.

C. Have reached 10% of the10 CFR 100 limits.

D. Have reached the 10 CFR 100 limits.

Question # 50 Revision: 0 KA #: 073000 A1.01 Tier 2 Group 1 Process Radiation Monitoring System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PRM system controls including: Radiation levels Importance: 3.2 / 3.5 CFR Number: 5.41(b)(11)

Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Fort Calhoun Objective:

STATE the two alarm setpoints for most monitors and EXPLAIN what each setpoint designates.

EXPLANATION:

"A" is the correct answer per the reference. "B" is the value for the Alarm setpoint. Choices "C" and "D" are related to other offsite radiation limits.

KA#: 073000 A1.01 Bank Ref #: 07-12-03 018 LP# / Objective: 0712-03 02.02 Exam Level: RO-11 Cognitive Level: LOW Source: BANK

Reference:

STM 33 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 051 The power supply to Raw Water Pump, AC-10D, is:

A. 480V Bus 1B3C B. 480V Bus 1B4C C. 4160V Bus 1A3 D. 4160V Bus 1A4 Question # 51 Revision: 0 KA #: 076000 K2.01 Tier 2 Group 1 Service Water System Knowledge of bus power supplies to the following: Service water Importance: 2.7 / 2.7 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN the automatic start features associated with the raw water pumps.

EXPLANATION:

Bus 1A4 provides power to AC-10D (choice D) The distractors are other vital buses at FCS.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 052 Given the following plant conditions:

x A plant transient caused the PORVs to cycle x The QUENCH TANK TEMP HI alarm was received on CB-1,2,3, A4 Following the transient, what actions should be taken to restore quench tank temperature to normal?

A. Supply fresh nitrogen to the quench tank while venting to containment atmosphere until the high temperature alarm clears.

B. Supply fresh nitrogen to the quench tank while venting to the vent header until the high temperature alarm clears.

C. Feed the quench tank with deaerated water while draining to a containment floor drain until the high temperature alarm clears.

D. Feed the quench tank with deaerated water while draining to the RCDT until the high temperature alarm clears.

Question # 52 Revision: 0 KA #: 007000 K4.01 Tier 2 Group 1 Pressurizer Relief Tank / Quench Tank System Knowledge of PRTS design feature(s) and/or interlock(s) which provide for the following: Quench tank cooling Importance: 2.6 / 2.9 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

LIST the major steps for proper operation of the quench tank per OI-RC-6.

EXPLANATION:

Choice "D" is the method used per the procedure. The distractors are other plausible methods.

KA#: 007000 K4.01 Bank Ref #: N/A LP# / Objective: 0711-20 03.04 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

OI-RC-6 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 053 Given the following plant conditions:

x The plant was operating at full power x Instrument air pressure was lost due to a header rupture inside containment x The CRS has entered AOP-17, "LOSS OF INSTRUMENT AIR" How will Instrument Air Containment Isolation Valves, PCV-1849A and PCV-1849B, be closed to isolate the break per the procedure?

A. They will both close automatically when instrument air pressure falls to 70 psig.

B. They will both be closed remotely by a Control Room Operator.

C. PCV-1849A will be closed locally by an Auxilary Building Operator which will cause PCV-1849B to close.

D. PCV-1849B will be closed locally by an Auxilary Building Operator which will cause PCV-1849A to close.

Question # 53 Revision:

KA #: 078000 K3.01 Tier 2 Group 1 Instrument Air System Knowledge of the effect that a loss or malfunction of the IAS will have on the following: Containment air system Importance: 3.1 / 3.4 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Explain the principles of normal operation of the Compressed Air System in terms of flow paths, major parameters, (temperature, pressure, flow, etc.), and control devices.

EXPLANATION:

These valves are operated from CB-10,11 in the control room. (choice B). Choice "A" would only be true with a CIAS. Choices "C" and "D" , Closing PCV-1849B will isolate air to the operator for PCV-1849A causing it to close.

KA#: 078000 K3.01 Bank Ref #: N/A LP# / Objective: 0711-07 01.04 Exam Level: RO-7 Cognitive Level: LOW Source: NEW

Reference:

AOP-17 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 054 The following conditions exist:

• A normal plant cooldown is in progress.

• PPLS has been blocked

• RCS pressure is 1615 psia and lowering

• Pressurizer pressure channel A/P-102 fails high.

Which one of the following will occur as a result of this sequence of events?

A. Only the PPLS Block 'A' circuit would be automatically reset.

B. Both PPLS circuits would be automatically reset but could be reblocked by operator action.

C. Both PPLS circuits would be automatically reset and could not be reblocked.

D. Both PPLS circuits would remain blocked.

Question # 54 Revision: 0 KA #: 013000 A4.01 Tier 2 Group 1 Engineered Safety Features Actuation System Ability to manually operate and/or monitor in the control room: ESFAS-initiated equipment which fails to actuate Importance: 3.5 / 4.8 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

KA#: 013000 A4.01 Bank Ref #: 07-12-14 011 LP# / Objective: 0711-07 01.05 Exam Level: RO-7 Cognitive Level: HIGH Source: NRC EXAM 1995

Reference:

Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 055 Which of the following will result in a Containment Isolation Actuation Signal (CIAS)?

A. SIAS B. SGIS C. CPHS D. CRHS Question # 55 Revision: 0 KA #: 103000 K4.06 Tier 2 Group 1 Containment System Knowledge of containment system design feature(s) and/or interlock(s) which provide for the following:

Containment isolation system Importance: 3.1 / 3.7 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN how each prime and backup actuation signal is developed.

EXPLANATION CIAS is caused by PPLS or CPHS, "C" is correct. "A" the same things that cause CIAS also cause SIAS, but SIAS does not cause CIAS. "B" and "D", SGIS isolates MSIV's and FW valves, CRHS causes VIAS.

KA#: 103000 K4.06 Bank Ref #:

LP# / Objective: 0712-14 01.04 Exam Level: RO-7 Cognitive Level: LOW Source: NRC 2005 EXAM

Reference:

STM 19 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 056 Given the following plant conditions:

x The plant is at full power x All CEAs are fully withdrawn x Rod Drive Clutches are being supplied from Instrument Buses A and C x Instrument Bus B is de-energized.

x The operator then inadvertently switches a clutch power supply from Instrument Bus A to Instrument Bus B How will CEA position be affected?

A. All trippable CEAs will fully insert.

B. One-half of the trippable CEAs will fully insert.

C. All CEAs will insert to the Power Dependent Insertion Limit.

D. All CEAs will remain fully withdrawn.

Question # 56 Revision:

KA #: 001000 K2.02 Tier 2 Group 2 Control Rod Drive System Knowledge of bus power supplies to the following: One-line diagram of power supply to trip breakers Importance: 3.6 / 3.7 CFR Number: 5.41(b)(6)

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

Fort Calhoun Objective:

Describe the interface/interaction between the CRDS and the following systems/components: Electrical Distribution System.

EXPLANATION:

The power going to the clutches is auctioneered, so even if there is no power from buses A and B, the clutches are still energized from bus C or D. Choice "D" is correct. The distractors are plausible if you don't understand how power is supplied to the CEDM clutches.

KA#: 001000 K2.02 Bank Ref #: 07-12-26 001 LP# / Objective: 0712-26 01.02A Exam Level: RO-6 Cognitive Level: HIGH Source: NRC 1995 EXAM

Reference:

STM 11 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 057 What enables the PORV LTOP feature where the PORV opening setpoints are automatically adjusted as RCS temperature lowers?

A. The feature is automatically enabled when pressurizer pressure falls below 1700 psia.

B. The feature is automatically enabled when RCS average temperature falls below 515°F.

C. The feature is enabled when the operator blocks SGLS.

D. The feature is enabled when the operator blocks PPLS.

Question # 57 Revision: 0 KA #: 002000 A3.03 Tier 2 Group 2 Reactor Coolant System Ability to monitor automatic operation of the RCS, including: Pressure, temperatures, and flows Importance: 4.4 / 4.6 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

GIVEN a copy of the Blocking PPLS floating step, EXPLAIN the steps necessary to initiate LTOP.

KA#: 002000 A3.03 Bank Ref #: 07-18-14 001 LP# / Objective: 0718-14 03.20 Exam Level: RO-7 Cognitive Level: LOW Source: NRC 1995 EXAM

Reference:

LP 0718.14 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 058 Given the following plant conditions:

x A slow plant load reduction is being performed x Pressurizer Level Control Channel, LRC-101Y has failed x Pressurizer Level Control Channel, LRC-101X is selected as the controlling channel and has inadvertently been placed in "AUTOMATIC" instead of "CASCADE" If no operator action is taken, how will pressurizer level control respond during the load reduction ?

A. Pressurizer level will be controlled to follow the reference level.

B. Pressurizer level will be controlled to stay above the reference level.

C. Pressurizer level will be controlled to stay below the reference level.

D. Pressurizer level will not be controlled.

Question # 58 Revision: 0 KA #: 011000 K6.04 Tier 2 Group 2 Pressurizer Level Control System Knowledge of the effect of a loss or malfunction on the following will have on the PZR LCS: Operation of PZR level controllers Importance: 3.1 / 3.1 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

EXPLAIN the interlocks and control functions associated with RCS Instrumentation.

EXPLANATION:

Programed pressurizer level goes from 60% to 48% during a power reduction. If the level controller is taken to automatic, it will control at the current level instead of the programmed level. This will be above the progranmmed level. Choice "B" is correct. The distractors are incorrect but plausible if you dont know how the system works.

KA#: 011000 K6.04 Bank Ref #: N/A LP# / Objective: 0711-20 04.04 Exam Level: RO-7 Cognitive Level: HIGH Source: NEW

Reference:

STM 36 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 059 Given the following plant conditions:

x A small break LOCA has occurred x A PPLS signal was generated and all systems performed as designed x RCS pressure indicates 991 psia and is stable x S/G pressure is being controlled at 900 psia by PCV-910.

x Subcooling indicated by the CET's is 38°F x Pressurizer level indicates 0%

x RVLMS level indicates 100%

x Two HPSI pumps and two LPSI pumps are running x Three charging pumps are running x Total HPSI flow indicates 415 gpm Tripping a __________ Pump would result in the largest reduction in subcooling margin as indicated by the CETs.

A. Charging B. Boric Acid C. HPSI D. LPSI.

Question # 59 Revision: 0 KA #: 017000 A1.01 Tier 2 Group 2 In-Core Temperature Monitor System Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ITM system controls including: Core exit temperature Importance: 3.7 / 3.9 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective:

EXPLAIN how the "Stop and Throttle" criteria is used to prevent reducing HPI flow when full HPI flow is required.

EXPLANATION:

Subcooling is being maintained by HPSI flow pressurizing the RCS. Thus tripping a HPSI pump will have the biggest affect on subcooling. (choice "C") The reduction in flow by tripping a charging pump would be smaller, The Boric Acid Pump is feeding the suction of the charging pump and the RCS pressure is above the LPSI pump shutoff head. The distractors are all incorrect.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: 017000 A1.01 Bank Ref #: N/A LP# / Objective: 0715-23 02.06 Exam Level: RO-5 Cognitive Level: HIGH Source: NEW

Reference:

EOP/AOP ATTACHMENT 2 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 060 Containment purge supply fans are not available. During a Containment Purge release using Containment Purge exhaust fan VA-32A, which damper is used to throttle the purge flow rate?

A. The Purge Exhaust Fan inlet damper (HCV-749)

B. The Purge Exhaust Fan outlet damper (YCV-747)

C. The outside Containment Purge exhaust isolation valve (HCV-742B)

D. The Purge Air Bypass Dilution damper (HCV-751)

Question # 60 Revision: 0 KA #: 029000 A4.01 Tier 2 Group 2 Containment Purge System Ability to manually operate and/or monitor in the control room: Containment purge flow rate Importance: 2.5 / 2.5 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

STATE the function of each major component of the Containment Purge System.

EXPLANATION:

The procedure directs that HCV-749 be used to throttle flow. The distracters are other dampers in the system.

KA#: 029000 A4.01 Bank Ref #: 07-14-04 LP# / Objective: 0714-04 01.04 Exam Level: RO-10 Cognitive Level: LOW Source: BANK

Reference:

OI-VA-1 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 061 Overload protection while using Spent Fuel Handling Machine, FH-12, to relocate spent fuel elements in the pool is provided by:

A. An interlock that stops vertical motion and cannot be overridden.

B. An interlock that stops vertical motion but can be overridden.

C. Only an indicator light that alerts the operator to stop vertical motion.

D. Only an indicator dial that is monitored by the operator to ensure overload limits are not exceeded.

Question # 61 Revision: 0 KA #: 034000 K4.03 Tier 2 Group 2 Fuel Handling Equipment System Knowledge of design feature(s) and/or interlock(s) which provide for the following: Overload protection Importance: 2.6 / 3.3 CFR Number: 5.41(b)(7)

Design, components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes,and automatic and manual features.

Fort Calhoun Objective:

Explain the function of the major components of the refueling machine and how interlocks prevent unsafe operation.

EXPLANATION:

"B" is correct, overload interlock will stop the hoist's vertical motion but can be overridden. The distractors are all incorrect but the hoist could have been designed that way.

KA#: 034000 K4.03 Bank Ref #: N/A LP# / Objective: 0711-13 01.02 Exam Level: RO-7 Cognitive Level: LOW Source: NEW

Reference:

STM 40 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 062 Given the following plant conditions:

x The reactor has tripped from full power as a result of a loss of all offsite power x Diesel Driven Auxiliary Feedwater Pump, FW-54, is tagged out of service x Diesel Generator, D-1, failed to start x Steam generator levels are currently 50% WR and lowering slowly x All safety functions, other than heat removal, are satisfied What action should be taken to establish heat removal?

A. Start AFW Pump, FW-6 B. Start AFW Pump, FW-10 C. Establish Once-through-Cooling D. Steam both S/Gs to establish conditions for Shutdown Cooling Question # 62 Revision: 0 KA #: 035000 A2.02 Tier 2 Group 2 Steam Generator System Ability to (a) predict the impacts of the following malfunctions or operations on the GS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Reactor trip/turbine trip Importance: 4.2 / 4.4 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective: EXPLAI the major strategy used to mitigate the consequences of a loss of all feedwater EXPLANATION:

With a loss of offsite power, main FW pumps are not available. FW-54 is out of service. There is no power to FW-6 because D/G D-1 did not start. FW-10 should be started (Choice "B"). "A" is incorrect, FW-6 has no power. "C" is incorrect, WR level is greater than 27%, "D" is incorrect because S/G feed is needed.

KA#: 035000 A2.02 Bank Ref #: 07-18-10 014 LP# / Objective: 0718-16 01.00 Exam Level: RO-10 Cognitive Level: HIGH Source: NRC EXAM 2001-2

Reference:

EOP-00 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 063 Given the following sequence of events:

x The reactor was stable at 12% power x The turbine tripped due to an overspeed test x Two steam dump valves failed open causing a power increase How will the plant respond?

A. Reactor power will rise and stabilize below 15%.

B. The reactor will trip when power exceeds 15%.

C. The reactor will trip when power exceeds 19.1%.

D. Reactor power will rise and stabilize above 19.1% power.

Question # 63 Revision: 0 KA #: 045000 K3.01 Tier 2 Group 2 Main Turbine Generator System Knowledge of the effect that a loss or malfunction of the MT/G system will have on the following:

Remainder of the plant Importance: 2.9 / 3.2 CFR Number: 5.41(b)(6)

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

Fort Calhoun Objective:

STATE the NSSS parameters and points that enable, disable and/or permit the following RPS trip functions:

EXPLANATION:

With the turbine tripped, the Reactor will trip on loss of load at 15% power. ("B") is correct.Two steam dump valves will remove more than 15% power thus "A" is incorrect. "C" is the variable high power trip setpoint but is above 15%. "D" would be correct if no trip setpoint were reached.

KA#: 045000 K3.01 Bank Ref #: 07-12-25 016 LP# / Objective: 0712-25 01.09 Exam Level: RO-6 Cognitive Level: HIGH Source: NRC EXAM 2001-1

Reference:

STM 38 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 064 Given the following plant conditions:

x A loss of normal feedwater has occurred due to a pipe rupture x Turbine Driven Auxiliary Feedwater Pump, FW-10, is supplying water to the steam generators x Level is lowering in the Emergency Feedwater Storage Tank Which one of the following pumps is used to supply makeup water to the Emergency Feedwater Storage Tank from the Condensate Storage Tank per AOP-30, "EMERGENCY FILL OF THE EMERGENCY FEEDWATER STORAGE TANK?"

A. Demineralized Water Pump, DW-40A B. Motor Driven AFW Pump, FW-6 C. Diesel Driven AFW Pump, FW-54 D. Diesel Driven Fire Pump, FP-1B Question # 64 Revision: 0 KA #: 056000 2.1.23 Tier 2 Group 2 Condensate System Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Importance: 4.3 / 4.4 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Use the Emergency Fill of EFWST Procedure to makeup to the tank if it goes below Technical Specification levels and normal makeup is not available.

EXPLANATION:

FW-54 is used to supply water to the EFWST from the CST. (C). Choices "A" and "D" can be used to makeup to the EFWST, but not from the CST. Choice "B" is incorrect, but plausible if you don't know the flow paths.

KA#: 056000 2.1.23 Bank Ref #: N/A LP# / Objective: 0717-30 01.00 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

AOP-30 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 065

__________ is used as a fire fighting agent in the Switchgear Rooms instead of water to prevent __________.

A. Halon, Flooding of the equipment below in room 19 B. Halon, Shorting of the electrical equipment in the Switchgear Rooms C. CO2 , Flooding of the equipment below in room 19 D. CO2 , Shorting of the electrical equipment in the Switchgear Rooms Question # 65 Revision:

KA #: 086000 K5.03 Tier 2 Group 2 Fire Protection System Knowledge of the operational implication of the following concepts as they apply to the Fire Protection System: Effect of water spray on electrical components Importance: 3.1 / 3.4 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Describe the major recovery actions of this AOP.

EXPLANATION:

Halon is used for fire protection in the swithgear rooms. (choice B) "A" is incorrect although room 19 is located below the switchgear rooms. "C" and "D" are incorrect because CO2 is not used in the switchgear room although it is used in the plant.

KA#: 086000 K5.03 Bank Ref #: N/A LP# / Objective: 0717-07 01.02 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

STM 21 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 066 In the Control Room, tan octagon labels or tan magnetic tags and Form FC-1290 are used to identify:

A. RCS Boric Acid Flowpath components during outages.

B. The HPSI Pumps that are aligned for autostart.

C. Preferred Raw Water Pumps and Heat Exchanger valves during all modes of operation.

D. Protected equipment during all modes of operation.

Question # 66 Revision: 0 KA #: 000000 2.1.31 Tier 3 Group 4 Generic Knowledges and Abilities Ability to locate control room switches, controls and indications and to determine that they correctly reflecting the desired plant lineup.

Importance: 4.6 / 4.3 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

RCS boration EXPLANATION:

"A" is correct per OPD-6-08. The distractors are all labels or tags used in the contrrol room, but not tan, octagon tags.

KA#: 000000 2.1.31 Bank Ref #: N/A LP# / Objective: 0707-42 10.06 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

OPD-6-08 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 067 Which one of the following methods is acceptable for verifying that the most current revision of a Operating Instruction (OI) procedure is being used?

A. Use Indexes on the Document Control Web page.

B. Contact the System Engineer for that latest revision.

C. Check Attachment 1 to SO-G-7, "PROCEDURE USE AND ADHERENCE."

D. Check with the Operations Procedure group Question # 67 Revision: 0 KA #: 000000 2.1.21 Tier 3 Group 4 Generic Knowledges and Abilities Ability to verify the controlled procedure copy.

Importance: 3.5 / 3.6 CFR Number: 5.41(b)(7)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:STATE the major sections of the Standing Orders.

EXPLANATION:

"A" is correct per the procedure. The distractors are all incorrect.

KA#: 000000 2.1.21 Bank Ref #: N/A LP# / Objective: 0717-17 01.02 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

STM 15 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 068 Given the following plant conditions:

x All CEAs are fully inserted and preparations are being made to perform a reactor startup by CEA withdrawal.

x According to the Estimated Critical Condition Calculation, boron concentration should be reduced by 250 ppm prior to the startup.

According to OP-2A,"PLANT STARTUP" which one of the following sequences of steps is acceptable?

A. Withdraw the non-trippable CEAs, dilute to the ECC boron concentration, withdraw the shutdown CEAs, withdraw the regulating CEAs.

B. Dilute to the ECC boron concentration, withdraw the shutdown CEAs, with draw the non-trippable CEAs, withdraw the regulating CEAs.

C. Withdraw the shutdown CEAs, withdraw the non-trippable CEAs, withdraw the regulating CEAs, dilute to the ECC boron concentration.

D. Withdraw the shutdown CEAs, dilute to the ECC boron concentration, withdraw the non-trippable CEAs, withdraw the regulating CEAs.

Question # 68 Revision: 0 KA #: 000000 2.2.01 Tier 3 Group 4 Generic Knowledges and Abilities Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

Importance: 4.5 / 4.4 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

Explain the operation of the Control Rod Drive System (CRDS).

EXPLANATION:

The reactivity management concept here is to not add positive reactivity without the ability to trip the reactor. ("D" is correct.) "A" and "B" would add positive reactivity without trip capability. "C" would not be a reactor startup by CEA withdrawal.

KA#: 000000 2.2.01 Bank Ref #: 07-12-26 017 LP# / Objective: 0712-26 01.00 Exam Level: RO-10 Cognitive Level: HIGH Source: NRC 1997 EXAM

Reference:

OP-2A Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 069 According to TDB-V.1.B, "ESTIMATED CRITICAL CONDITIONS WORKSHEET,"

which one of the following situations may require adjustment to the required boron concentration to compensate for changes in B-10 concentration?

A. The RCS has been cooled down to less than 210°F.

B. The RCS has been drained to mid-loop and refilled.

C. The reactor has been shutdown for more than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

D. "Acid Reducing" conditions have been established in the RCS.

Question # 69 Revision:

KA #: 000000 2.1.43 Tier 3 Group 4 Generic Knowledges and Abilities Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

Importance: 4.1 / 4.3 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

DESCRIBE and USE the shutdown margin worksheets.

EXPLANATION:

Boron depletion, lowering B-10 content is RCS boron, occurs during operation. An operation that adds a lot of newly batched boric acid changes the content. "B" is correct. Distractors "A" and "C" both affect shurdown margin. "D" is done as part of a shutdown for chemistry control.

KA#: 000000 2.1.43 Bank Ref #: N/A LP# / Objective: 0705-09 05.00 Exam Level: RO-10 Cognitive Level: HIGH Source: NEW

Reference:

TDB-V.1.B Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 070 How is the installation of a temporary modification [TM] on a system annotated on the controlled P&IDs in the Control Room (copy #14) and the OCC (copy #1)?

A. The associated P&ID drawings are marked with green dots bearing the temporary modification number.

B. Magenta "Post-It" notes listing the temporary modification number are attached to the associated P&ID drawings.

C. The associated P&ID drawings are replaced with temporary drawings that reflect the temporary modification.

D. The associated P&ID drawings are marked up in blue to indicate the location of the temporary modification.

Question # 70 Revision: 0 KA #: 000000 2.2.41 Tier 3 Group 4 Generic Knowledges and Abilities Ability to obtain and interpret station electrical and mechanical drawings.

Importance: 3.5 / 3.9 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

STATE the major sections of the Standing Orders.

EXPLANATION:

"A" is the method that is used. The distractors are other viable methods.

KA#: 000000 2.2.41 Bank Ref #: ADM-OPS 034 LP# / Objective: 0762-01 01.00 Exam Level: RO-10 Cognitive Level: LOW Source: NRC EXAM 2002

Reference:

SO-O-25 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 071 Given the following conditions:

x The plant is operating at full power x You are exiting a posted contaminated area x A step-off pad has been provided for removal of protective apparel x There are no friskers located by the stepoff pad but there is one 30 feet away Which one of the following actions should you take?

A. Contact Radiation Protection prior to stepping on the step off pad.

B. Contact Radiation Protection after stepping on the step off pad.

C. Perform a hand and foot frisk at the nearest frisker location.

D. Proceed directly to the RCA exit and use a PCM to check for contamination.

Question # 71 Revision: 0 KA #: 000000 2.3.05 Tier 3 Group 4 Generic Knowledges and Abilities Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Importance: 2.9 / 2.9 CFR Number: 5.41(b)(12)

Radiological safety principles and procedures.

Fort Calhoun Objective:

EXPLAIN the proper method of frisking, what background countrate is acceptable, and when an individual is considered contaminated.

EXPLANATION:

"C" is correct per the reference. "A" and "B" are incorrect because not every stepoff pad has a frisker. "D" is incorrect during power operation.

KA#: 000000 2.3.05 Bank Ref #: N/A LP# / Objective: 1924-03 01.17 Exam Level: RO-12 Cognitive Level: LOW Source: NEW

Reference:

SO-G-101 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 072 The administrative exposure limit from occupational sources at FCS is:

A. 0.3 Rem/year TEDE B. 1.0 Rem/year TEDE C. 4.0 Rem/year TEDE D. 5.0 Rem/Year TEDE Question # 72 Revision: 0 KA #: 000000 2.3.13 Tier 3 Group 4 Generic Knowledges and Abilities Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Importance: 3.4 / 3.8 CFR Number: 5.41(b)(11)

Purpose and operation of radiation monitoring systems, including alarms and survey equipment.

Fort Calhoun Objective: State the FCS radiation exposure limits.

EXPLANATION:

"B" is correct per SO-G-101. "A" is the limit for pregnant workers. "C" is the FCS limit for all occupational sources. "D" is the regulatory limit.

KA#: 000000 2.3.13 Bank Ref #:

LP# / Objective: 0715-33 01.03 Exam Level: RO-11 Cognitive Level: LOW Source: NEW

Reference:

SO-G-101 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 073 A plant cooldown is being performed using OP-3A, "PLANT SHUTDOWN." RCS Temperature is 450°F.

What action should be taken If an event should occur and specific AOP or ARP guidance can not be located?

A. Enter EOP-00 and perform Standard Post Trip Actions.

B. Reference EOP-00 and the appropriate EOP and follow procedural guidance verbatim.

C. Reference EOP-00 and the appropriate EOP and selectively follow procedural guidance.

D. Contact the TSC and wait for specific guidance.

Question # 73 Revision: 0 KA #: 000000 2.4.09 Tier 3 Group 4 Generic Knowledges and Abilities Knowledge of low power / shutdown implications in accident (e.g. Loss of coolant accident loss of residual heat removal) mitigation strategies.

Importance: 3.8 / 4.2 CFR Number: 5.41(b)(5)

Facility operating characteristics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of load changes, and operating limitations and reasons.

Fort Calhoun Objective: STATE the major sections of the Standing Orders.

EXPLANATION: SO-O-1 states that when T-cold is less than 525°F and the plant is not on shutdown cooling, an event occurs and specific AOP or ARP guidance can not be found, refer to EOP-00 and selectively follow procedural guidance. "C" is correct. "A" is incorrect, EOP-00 is not entered. "B" is incorrect, verbatim compliance is not required. "D" is incorrect, TSC guidance is not required.

KA#: 000000 2.4.09 Bank Ref #: N/A LP# / Objective: 0767-05 02.00 Exam Level: RO-5 Cognitive Level: HIGH Source: NEW

Reference:

SO-O-1 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 074 Which of the following programs is used to define risk profiles and assign risk assessment "colors" as conditions chage during outages" per SO-O 21, "SHUTDOWN OPERATIONS PROTECTION PLAN?"

A. EOOS B. ORAM C. GARDEL D. EAGLE Question # 74 Revision: 0 KA #: 000000 2.2.18 Tier 3 Group 4 Generic Knowledges and Abilities Knowledge of the process for managing maintenance activities during shutdown operations, such as risk assessments, work prioritization, etc.

Importance: 2.6 / 3.9 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

DESCRIBE the significant risk contributing scenarios identified by industry experience and PRA studies.

EXPLANATION:

ORAM is used to assign risk colors during outages. (B). "A" is used for risk colors during operation. "C" and "D" are computer programs used for core monitoring and offsite dose projection.

KA#: 000000 2.2.18 Bank Ref #: N/A LP# / Objective: 0707-42 02.01 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

SO-O-21 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 075 The Emergency Plan has just been entered due to a plant event. Who is the Emergency Director prior to activation of the Technical Support Center and the Emergency Offsite Facility?

A. The Shift Manager.

B. The Control Room Supervisor.

C. The Shift Technical Advisor.

D. The Work Week Manager Question # 75 Revision: 0 KA #: 000000 2.4.37 Tier 3 Group 4 Generic Knowledges and Abilities Knowledge of the lines of authority during implementation of the emergency plan.

Importance: 3 / 4.1 CFR Number: 5.41(b)(10)

Administrative, normal, abnormal, and emergency operating procedures for the facility.

Fort Calhoun Objective:

The person who completes this topic will be able to IDENTIFY specifices of the organization and methods of operation of the Emergency Response Organization.

EXPLANATION:

The Shift Manger initially serves as the Emergency Director havig the Command and Control position. (A) the other choices are incorrect.

KA#: 000000 2.4.37 Bank Ref #: N/A LP# / Objective: 1070-001 1.0 Exam Level: RO-10 Cognitive Level: LOW Source: NEW

Reference:

EPIP-OSC-2 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 076 Given the following plant conditions:

x The plant has tripped from full power due to a High Pressurizer Pressure Trip x Standard Post-Trip actions were taken and EOP-01, "REACTOR TRIP RECOVERY," was entered x Several Minutes later PPLS actuated x All Engineered Safety Features operated as designed x Pressurizer Pressure is now 931 psia x The ATCO has tripped all Reactor Coolant Pumps x Pressurizer Level is 100% (solid) x Representative CET Temperature is 536°F x PORV, PCV-102-2, is discovered to be open Which one of the following actions should be taken?

A. Reenter EOP-00, "STANDARD POST TRIP ACTIONS," then enter EOP-03, LOSS OF COOLANT ACCIDENT,and close HCV-151, the block valve for PCV-102-2.

B. Stay in EOP-01, REACTOR TRIP RECOVERY," and reestablish letdown using EOP/AOP Attachment 23, "RESTORATION OF LETDOWN."

C. Reenter EOP-00, "STANDARD POST TRIP ACTIONS," then enter EOP-03, LOSS OF COOLANT ACCIDENT, and begin a plant cooldown.

D. Enter EOP-20, FUNCTIONAL RECOVERY PROCEDURE, Implement EOP/AOP Floating Step A, "HPSI STOP AND THROTTLE CRITERIA."

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 76 Revision: 0 KA #: 000008 2.4.20 Tier 1 Group 1 Pressurizer Vapor Space Accident Knowledge of operational implications of EOP warnings, cautions, and notes.

Importance: 3.8 / 4.3 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

DEMONSTRATE the knowledge required to use EOP-03, Loss of Coolant Accident (LOCA), to mitigate the consequences of a LOCA.

EXPLANATION:

EOP-01, directs you to EOP-00, Diagnostic Actions. A caution in EOP-03 states to not close the PORV block valves with the pressurizer full but instead begin a plant cooldown. ("C" is correct) "A" is incorrect because it closes the block valve. Choices "B" and "D" can help control pressure, but are not procedurely directed under the stem conditions.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 077 Given the following plant conditions:

x The Reactor has been shutdown for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> x Shutdown Cooling was in service using LPSI Pump, SI-1A x The RCS Temperature was at 190°F and lowering slowly x The RCS Pressure is 15 psia x Containment Pressure is 15 psia x Pressurizer Level is 50%

x The Pressurizer Manway has been removed x The SIRWT is full x The Refueling Cavity has not been filled x All Containment Spray Pumps, HPSI Pumps and Charging Pumps are available x AFW Pumps FW-6 and FW-54 are available x LPSI Pump, SI-1A, tripped 1 minute ago x LPSI Pump, SI-1B will not start Which one of the following alternate cooling strategies should be implemented per AOP-19, "LOSS OF SHUTDOWN COOLING"?

A. Establish a minimum of 55 gpm auxiliary feedwater flow and Implement Attachment F, "Alternate Decay Heat Removal using Steam Generators."

B. Establish a minimum of 575 gpm HPSI flow using Attachment N, "Shutdown Cooling via the HPSI header."

C. Establish a minimum of 120 gpm Charging Flow using Attachment E, "Alternate Decay Heat Removal by Boiling."

D. Establish a minimum of 1500 gpm Shutdown Cooling flow using Attachment Q, "Alternate Shutdown Cooling Utilizing Containment Spray Pumps."

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 77 Revision: 0 KA #: 000025 AA2.05 Tier 1 Group 1 Loss of Residual Heat Removal System Ability to determine and interpret the following as they apply to the Loss of Residual Heat Removal System: Limitations on LPI flow and temperature rates of change Importance: 3.1 / 3.5 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Use the Loss of Shutdown Cooling Procedure to mitigate the consequences of a loss of cooling to the Reactor Coolant System.

EXPLANATION: Pressure boundary not intact, RV head on, SDC discharge not available, HPSI discharge is available, Sufficient injectionis available, Implement N, 575 gpm from below. "B" is the only correct answer.

KA#: 000025 AA2.05 Bank Ref #: N/A LP# / Objective: 0717-19 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

AOP-19 Handout: AOP-19 ATT D CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 078 Given the following plant conditions:

x The Reactor has tripped x EOP-00, "STANDARD POST TRIP ACTIONS," have been completed x EOP-01, "REACTOR TRIP RECOVERY," has been entered x The STA was in the process of performing the EOP-01, Safety Function Status Check when the BATTERY CHARGER #1 TROUBLE and the DC BUS #1 LOW VOLTAGE annunciators alarmed x Voltage on DC Bus #1 indicates 120 Volts x The STA reports that the Safety Function Status Check is unacceptable because DC Bus #1 is not being powered by a Battery Charger.

What action should be taken?

A. Stay in EOP-01. Follow the ARP guidances to power DC Bus #1 using Battery Charger #3 per OI-EE-3, "125 VDC SYSTEM NORMAL OPERATION."

B. Stay in EOP-01, Implement AOP-16,Section VIII, "LOSS OF DC BUS 1."

C. Enter EOP-20, "FUNCTIONAL RECOVERY PROCEDURE," and perform actions for Safety Function MVA-DC.

D. Return to EOP-00, "DIAGNOSTIC ACTIONS" section, then enter EOP-20 and perform actions for Safety Function MVA-DC.

Question # 78 Revision: 0 KA #: CE-E02 EA2.01 Tier 1 Group 1 Reactor Trip Recovery Ability to determine and interpret the following as they apply to the (Reactor Trip Recovery) Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Importance: 2.7 / 3.7 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

GIVEN a copy of the Safety Function Status Check acceptance criteria, DETERMINE the parameters used to confirm an uncomplicated Reactor Trip.

EXPLANATION:

EOP-01 directs a return to EOP-00 Diagnostic Actions if a Safety Function Status Check is not acceptable. Diagnostic actions will then direct entry into EOP-20. "D" is correct. The distractors are all incorrect because they do not return to EOP-00.

EOP-00 page 33 flow chart CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: CE-E02 EA2.01 Bank Ref #: N/A LP# / Objective: 0718-11 01.03 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

EOP-01 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 079 Given the following plant conditions:

x The Reactor tripped from full power x Multiple CEAs failed to insert on the trip x All applicable EOP-00, "STANDARD POST TRIP ACTIONS" have been taken and WR NI Power is indicating 7% and steady What action should be taken at the completion of EOP-00?

A. Enter AOP-02, "CEA AND CONTROL SYSTEM MALFUNCTIONS," and insert CEAs using the Rod Drop Test Switches.

B. Enter AOP-03, "EMERGENCY BORATION" and borate from the Safety Injection and Refueling Water Storage Tank.

C. Enter EOP-20, "FUNCTIONAL RECOVERY PROCEDURE," and Insert the Group "N" CEAs manually.

D. Enter EOP-20, "FUNCTIONAL RECOVERY PROCEDURE," and Trip the AI-57 Power Supply Breakers on AI-40A/B/C/D.

Question # 79 Revision: 0 KA #: 000029 2.4.11 Tier 1 Group 1 Anticipated Transient Without Scram (ATWS)

Knowledge of abnormal condition procedures.

Importance: 4 / 4.2 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

USE the Functional Recovery Procedure (EOP-20) to bring the reactor, Reactor Coolant System and containment to a safe and stable condition.

EXPLANATION:

Guidance for tripping the AI-57 power supply breakers on AI-40A/B/C/D is only found in EOP-20. "D" is correct. "C" inserting group "N" CEAs is helpful, but not directed by EOP-20. "A" and "B" are incorrect because there is no transition from EOP-00 to these procedures.

KA#: 000029 2.4.11 Bank Ref #: N/A LP# / Objective: 0718-18 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

EOP-20 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 080 Given the following plant conditions:

x The Reactor is operating at full power x An Urgent High Voltage ERFCS alarm was received for bus 1A3 x Bus 1A3 voltage has indicated 4395 V for one hour x 161 KV Grid Voltage has indicated 170 KV for one hour x System Operations has been contacted and requested to lower grid voltage What additional action should be taken and why?

A. Select MANUAL voltage control using AOP-27, "GENERATOR MALFUNCTIONS,"

to hold the generator field voltage constant.

B. Start a second Stator Water Cooling Pump using AOP-27, "GENERATOR MALFUNCTIONS," to provide additional cooling to the main Generator.

C. Start redundant safety related loads powered from bus 1A4 and shutdown safety related loads powered from bus 1A3 using AOP-31, "161 KV GRID MALFUNCTIONS," to prevent damage to equipment due to overvoltage.

D. Start additional loads powered by bus 1A3 using AOP-31, "161 KV GRID MALFUNCTIONS," to reduce voltage on bus 1A3.

Question # 80 Revision: 0 KA #: 000077 2.4.18 Tier 1 Group 1 Generator Voltage and Electric Grid Disturbances Knowledge of the specific bases for EOPs.

Importance: 3.3 / 4 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Describe the major recovery actions of this AOP.

EXPLANATION:

The stem conditions meet the entry conditions for AOP-31, which directs additional loads be started to lower bus voltage. ("D" is correct) Choices "A" and "B" enter the wrong procedure. Choice "C" shuts down loads on the bus and is incorrect.

KA#: 000077 2.4.18 Bank Ref #: N/A LP# / Objective: 0717-31 01.03 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

TBD-AOP-31 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 081 Given the following plant conditions:

x The plant is operating at full power x Instrument Air Compressor, CA-1C, is out of service for maintenance x The "INSTRUMENT AIR PRESS LO" and "PLANT AIR PRESS LO" annunciators are in alarm x AOP-17, "LOSS OF INSTRUMENT AIR," has been entered and actions taken x Instrument air pressure has stabilized at 82 psig with CA-1A and CA-1B fully loaded Which one of the following actions should be taken per AOP-17?

A. Trip the Reactor and enter EOP-00, "STANDARD POST TRIP ACTIONS."

B. Evaluate the need to shutdown the Reactor per AOP-05, "EMERGENCY SHUTDOWN."

C. Declare a "Notice of Unusual Event." per EPIP-OSC-1, "EMERGENCY CLASSIFICATION,"

D. Exit AOP-17 and continue full power operation.

Question # 81 Revision: 0 KA #: 000065 AA2.05 Tier 1 Group 1 Loss of Instrument Air Ability to determine and interpret the following as they apply to the Loss of Instrument Air: When to commence plant shutdown if instrument air pressure is decreasing Importance: 3.4 / 4.1 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Use the Loss of Instrument Air Procedure to mitigate the consequences of a partial or complete loss of instrument air.

EXPLANATION:

With the pressure above 50 psig but below 98 psig, AOP-17 directs evaluation of the need to shutdown.

Choice "B" is correct. "A" is incorrect because pressure is greater than 50 psig. "D" is incorrect because pressure is less than 98 psig. "C" is incorrect because this is not addressed in the EPIPs.

KA#: 000065 AA2.05 Bank Ref #: N/A LP# / Objective: 0717-17 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

AOP-17 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 082 Given the following plant conditions:

x The plant was operating at full power x A large LOCA occurred 5.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> ago x All ESF equipment operated as designed x RAS occurred 3.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> ago x RCS pressure is 17 psia x Containment pressure is 2 psig x Only one HPSI Pump is available x RVLMS indicates 21%

Which one of the following actions is required per EOP-03?

A. Implement EOP/AOP Attachment 9, "SIMULTANEOUS HOT AND COLD LEG INJECTION," within 30 minutes.

B. Implement EOP/AOP Attachment 9, "SIMULTANEOUS HOT AND COLD LEG INJECTION," between 2.0 and 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> from now.

C. Implement EOP/AOP Attachment 11, "ALTERNATE HOT LEG INJECTION," within 30 minutes.

D. Implement EOP/AOP Attachment 11, "ALTERNATE HOT LEG INJECTION,"

between 2.0 and 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> from now.

Question # 82 Revision: 0 KA #: 000074 2.4.49 Tier 1 Group 2 Inadequate Core Cooling Ability to perform without reference to procedures those actions that require immediate operation of system components and controls.

Importance: 4.6 / 4.4 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

EXPLAIN the problems associated with boron precipitation for a cold leg break, what actions are taken to minimize it and why it is not a problem for a hot leg break.

EXPLANATION:

With the conditions in the stem, EOP/AOP Attachment 11 must be implemented immediately. Choice "C" is correct. Choice "D" is incorrect because it addresses 5.5 to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after RAS. Choices "A" and "B" are incorrect because Attachment 11 is used when only one HPSI pump is available.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 083 Given the following plant conditions:

x The plant is operating at Full Power x Condenser Offgas Radiation Monitor,RM-057, is in alert x Blowdown Radiation Monitor, RM-054B, indicates rising counts x Charging flow is 40 gpm x Letdown flow is 35 gpm x Pressurizer pressure and level are steady x Reactor Coolant Pump Seal parameters are normal x AOP-22, "REACTOR COOLANT LEAK," Attachment B, "Primary to Secondary Leak Rate Actions" has been entered x The Shift Chemist has confirmed the primary to secondary leakrate to be the same as indicated by primary plant parameters Which one of the following actions should be taken per AOP-22, Attachment B?

A. Maintain the current plant conditions. Continue to monitor for increased primary to secondary leakage.

B. Commence a 3% per hour power reduction to Mode 3 per OP-4, LOAD CHANGE AND NORMAL POWER OPERATION. When Mode 3 is entered, implement EOP-00, "STANDARD POST TRIP ACTIONS" C. Commence a power reduction to Mode 3 per AOP-05, "EMERGENCY PLANT SHUTDOWN." When Mode 3 is entered, implement EOP-00, "STANDARD POST TRIP ACTIONS" D. Trip the Reactor and enter EOP-00, "STANDARD POST TRIP ACTIONS" then transition to EOP-04, "STEAM GENERATOR TUBE RUPTURE."

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 83 Revision: 0 KA #: 000037 AA2.04 Tier 1 Group 2 Steam Generator Tube Leak Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak:

Comparison of RCS fluid inputs and outputs, to detect leaks Importance: 3.4 / 3.7 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Describe the Technical Specification LCO that is challenged by a leak in the Reactor Coolant System.

EXPLANATION:

The charging/letdown mismatch indicates primary to secondary leakage is 1 gpm and an AOP-05 shutdown is warranted by AOP-22, "C" is correct. "A" and "B" would be correct if the leak rate was smaller. "D" would be correct if the leak rate were greater than 40 gpm.

KA#: 000037 AA2.04 Bank Ref #: N/A LP# / Objective: 0717-22 01.06 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

AOP-22 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 084 Given the following plant conditions:

x The "CONDENSER VACUUM STANDBY PUMP RUNNING" annunciator is in alarm x All three condenser evacuation pumps are running x Condenser vacuum is 26.5" and steady x Generator load is 470 MWe x Water is visible in the separator tank level gage x The Circulating water system is operating normally x There is no water in, ST-13, the Steam Packing Exhauster Drain Trap x Vacuum Breaker, VD-200, is closed x A visible level is being maintained in, VD-4, Drip and Drain tank x Condensate Storage Tank, DW-48, Level is 90%

What action should be taken after entering the ARP based on these indications?

A. Initiate a plant shutdown using AOP-05, EMERGENCY SHUTDOWN." Ensure the isolation valves for HCV-1040, "Atmospheric Steam Dump Valve," are open.

B. Trip the Reactor and turbine and enter EOP-00, "STANDARD POST TRIP ACTIONS.

C. Initiate continuous feed and bleed of the Condenser Evacuation Separator Tanks using OI-CE-1, "CONDENSER EVACUATION SYSTEM NORMAL OPERATION."

D. Isolate the steam packing exhauster drain trap using AOP-26, "TURBINE MALFUNCTIONS."

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 84 Revision: 0 KA #: 000051 AA2.01 Tier 1 Group 2 Loss of Condenser Vacuum Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum: Cause for low vacuum condition Importance: 2.4 / 2.7 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Describe the major recovery actions of this AOP.

EXPLANATION:

AOP-26 directs isolation of the steam packing exhauster drain trap if no water is visible. Choice "D" is correct. "A" would be correct if vacuum was less than 25". "B" would be correct if vacuum was less than 23.85" Choice "C" is plausible becase it involves the condenser evacuation system.

KA#: 000051 AA2.01 Bank Ref #: 07-17-26 LP# / Objective: 0717-26 01.03 Exam Level: SRO-5 Cognitive Level: HIGH Source: NRC 2004 EXAM

Reference:

AOP-26 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 085 Given the following plant conditions:

x A Loss of Coolant Accident occurred 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> ago x DC Bus 1 was lost 5 minutes after the LOCA x EOP-20, "FUNCTIONAL RECOVERY PROCEDURE," was entered x DC Control power for appropriate equipment was transferred to DC Bus 2 x RAS has just occurred x HPSI pumps, SI-2A and SI-2B are operating x No LPSI or Containment Spray Pumps are operating x Pressurizer pressure is 50 psia x HPSI flow indication was fluctuating x HPSI Discharge header pressure was fluctuating x HPSI Pump performance improved after reducing flow to 50 gpm per pump What action should be taken next?

A. Maintain 50 gpm injection per pump in accordance with EOP-20 B. Start HPSI pump SI-2C and establish a total HPSI flow of 150 gpm in accordance with EOP-20.

C. Slowly increase HPSI flow to the minimum flow required by EOP/AOP Attachment 3, "Safety Injection Flow vs. Pressurizer Pressure."

D. Slowly increase HPSI flow to the minimum flow required by EOP/AOP Attachment 26, "Total SI Pump Flow to Match Decay Heat vs. Time After Trip."

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 85 Revision: 0 KA #: CE-E09 2.4.47 Tier 1 Group 2 Functional Recovery Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

Importance: 4.2 / 4.2 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Given the Resource Assessment Trees, basically DESCRIBE the Method, Path and Acceptance Criteria for each success path.

EXPLANATION:

With reduced suction flow, EOP-20 directs that flow be slowly increased to minimum flow in Attachemnt

26. (Choice "D") Choice "C" uses the wrong attachment. "B" is incorrect SI-2A and SI-2C should not be operarted together. Chice "A" would be correct if HPSI pump performance had not improved.

KA#: CE-E09 2.4.47 Bank Ref #: N/A LP# / Objective: 0718-18 01.05 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

EOP-20 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 086 Given the following plant conditions:

x The Plant has just tripped from full power x All plant parameters were normal prior to the trip x PPLS, SIAS, CIAS, VIAS actuations have occurred x Relays 86A/PPLS and 86A1/PPLS and associated actuation relays tripped x Relays 86B/PPLS and 86B1/PPLS and associated actuation relays are in normal position x EOP-00, "STANDARD POST TRIP ACTIONS" was entered and performed x All plant parameter values and trends are acceptable per EOP-00 What actions should be taken after completion of EOP-00?

A. Enter EOP-01, "REACTOR TRIP RECOVERY," then enter AOP-23, "RESET OF ENGINEERED SAFEGUARDS," to recover from the inadvertent PPLS actuation.

B. Enter AOP-23, "RESET OF ENGINEERED SAFEGUARDS," to recover from the inadvertent PPLS actuation then enter EOP-01, "REACTOR TRIP RECOVERY."

C. Perform EOP/AOP Floating Step H "RESET OF ENGINEERED SAFEGUARDS," to recover from the inadvertent PPLS actuation then Enter EOP-01, "REACTOR TRIP RECOVERY."

D. Enter EOP-01, "REACTOR TRIP RECOVERY," then perform EOP/AOP Floating Step H "RESET OF ENGINEERED SAFEGUARDS," to recover from the inadvertent PPLS actuation.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 86 Revision: 0 KA #: 006000 A2.13 Tier 2 Group 1 Emergency Core Cooling System Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadvertent SIS actuation Importance: 3.9 / 4.2 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Use the Reset of Engineered Safeguards Procedure to mitigate the consequences of an inadvertent safeguards actuation.

EXPLANATION:

EOP-00 sends you to EOP-01 in this situation. AOP-23 is used to recover from an inadvertent ESF actuation. "A" is correct. "B" is incorrect because you do not enter AOP-23 before entering EOP-01.

"C" and "D" are incorrect because Floating Step H does not address inadvertent safeguards actuation.

KA#: 006000 A2.13 Bank Ref #: N/A LP# / Objective: 0717-23 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

AOP-23 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 087 Given the following plant conditions:

x The plant was operating at full power when a LOCA occurred x The following actuations were received: PPLS, CPHS, CSAS, SIAS, CIAS and VIAS x EOP-03, "LOSS OF COOLANT ACCIDENT," was entered after performing EOP-00, "STANDARD POST TRIP ACTIONS" x 2 HPSI pumps, 2 LPSI pumps and 3 charging pumps are operating x All Reactor Coolant Pumps have been tripped x SIRWT Level is 122 inches x No Containment Spray pumps are operating x RCS Pressure is 500 psia and lowering slowly x Steam Generator pressures are 800 psia x Containment Pressure is 28 psig and rising slowly x SI flowrate is below the flow required by EOP/AOP Attachment 3, "Safety Injection Flow vs. Pressurizer Pressure" x Both Reactor Vessel Level Channels indicate 29%

x CETs indicate 467°F What actions should be taken to mitigate this situation?

A. Stay in EOP-03 and start a third HPSI pump.

B. Stay in EOP-03 and start a plant cooldown.

C. Enter EOP-20, "FUNCTIONAL RECOVERY PROCEDURE", Implement Success Path IC-2 and open PORVs.

D. Enter EOP-20, "FUNCTIONAL RECOVERY PROCEDURE", Implement Success Path CI and start two Containment Spray Pumps.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 87 Revision: 0 KA #: 013000 A2.01 Tier 2 Group 1 Engineered Safety Features Actuation System Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based Ability on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations; LOCA Importance: 4.6 / 4.8 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Given the Resource Assessment Trees, basically DESCRIBE the Method, Path and Acceptance Criteria for each success path.

EXPLANATION:

According to EOP-03, If safety injection flow is inadequate per EOP/AOP attachment 2, then all idle HPSI pumps should be started. "A" is correct. "B" is later in the procedure. "C" and "D" are incorrect because EOP-20 entry is not required.

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 088 Given the following plant conditions:

x The plant is operating at full power x Pressurizer Pressure and Level are lowering x All three charging pumps are running x Letdown flow is 26 gpm x The "RM-057 CONDENSER OFF GAS HIGH RADIATION" annunciator is in alarm x The "RM-054A STM GEN "A" BLWD HIGH RADIATION" annunciator is also in alarm What is the appropriate procedural guidance for placing MAIN STEAM LINE RADIATION MONITOR, RM-064, in service?

A. RM-064 will be placed in service on both steam generators using EOP-00, "STANDARD POST TRIP ACTIONS."

B. RM-064 will be placed in service on both steam generators using EOP-04, "STEAM GENERATOR TUBE RUPTURE."

C. RM-064 will be placed in service on the "A" steam generator only using the Annunciator Response Procedure for the "RM-057 CONDENSER OFF GAS HIGH RADIATION" alarm.

D. RM-064 will be placed in service on the "A" steam generator only using AOP-22, "REACTOR COOLANT LEAK," Attachment B, "Primary to Secondary Leak Rate Actions."

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 88 Revision: 0 KA #: 039000 A2.03 Tier 2 Group 1 Main and Reheat Steam System Ability to (a) predict the impacts of the following malfunctions or operations on the MRSS; and (b) based on predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Indications and alarms for main steam and area radiation monitors (during SGTR)

Importance: 3.4 / 3.7 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

EXPLAIN the major strategy used to mitigate the consequences of a SGTR.

EXPLANATION: Guidance for placing RM-064 into service is found in EOP-04, ARPs and AOP-22, but not EOP-00. "A" is incorrect. "B" is correct. "C" and "D" are incorrect because RM-064 is placed into operation for both S/Gs. (Auto alternates between the steam generators)

KA#: 039000 A2.03 Bank Ref #: N/A LP# / Objective: 0718-14 01.01 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

EOP-04 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 089 Given the following plant conditions:

x The Reactor Tripped Due to a loss of Loss of Offsite Power x EOP-02, "LOSS OF OFFSITE POWER/LOSS OF FORCED CIRCULATION."

was entered following EOP-00, STANDARD POST TRIP ACTIONS x Turbine Driven Auxiliary Feedwater Pump, FW-10, is providing water to the steam generators x The EFWST LEVEL LO APPROACHING TECH SPEC LIMIT annunciator is in alarm x The EFWST LEVEL LO TECH SPEC LIMIT annunciator is NOT in alarm x The EFWST LEVEL LO-LO annunciator is NOT in alarm x EFWST level indicates 89%

What action should be taken in response to these alarm conditions?

A. Implement EOP-AOP floating step J, "EMERGENCY FEEDWATER STORAGE TANK INVENTORY," at this time.

B. Implement AOP-30, "EMERGENCY FILL OF THE EMERGENCY FEEDWATER STORAGE TANK," at this time.

C. Implement EOP-AOP floating step J, "EMERGENCY FEEDWATER STORAGE TANK INVENTORY," when the EFWST LEVEL LO-LO alarm is received.

D. Implement AOP-30, "EMERGENCY FILL OF THE EMERGENCY FEEDWATER STORAGE TANK," when the EFWST LEVEL LO-LO alarm is received.

Question # 89 Revision: 0 KA #: 061000 2.4.46 Tier 2 Group 1 Auxiliary / Emergency Feedwater System Ability to verify that the alarms are consistent with the plant conditions.

Importance: 4.2 / 4.2 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

STATE from memory the sources of makeup water to replenish the EFWST as described in the EFWST Inventory floating step and AOP-30.

EXPLANATION:

Entry condition is met for floating step ("A" is correct), Not met for AOP-30 ("B" is incorrect) LO-LO level alarm is after entry conditions ("C" and "D" are incorrect)

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL KA#: 061000 2.4.46 Bank Ref #: N/A LP# / Objective: 0718-12 03.08 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

AOP-30 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 090 Given the following plant conditions:

x The plant is operating at full power x An I&C Technician has informed you that Pressure Transmitter, B/PT-102 can not be calibrated because it has failed low.

x The Channel "B" High Pressurizer Pressure and TM/LP trip units have been bypassed What additional actions are required by Technical Specifications?

A. 86B/PPLS must be bypassed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />..

B. 86A/PPLS and 86B/PPLS must be bypassed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

C. 86B/PPLS and 86B1/PPLS must be bypassed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. 86A/PPLS, 86B/PPLS, 86A1/PPLS and 86B1/PPLS must be bypassed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

Question # 90 Revision: 0 KA #: 012000 2.1.32 Tier 2 Group 1 Reactor Protection System Ability to explain and apply system limits and precautions.

Importance: 3.8 / 4 CFR Number: 5.43(b)(2)

Facility operating limitations in the technical specifications and their bases.

Fort Calhoun Objective:

EXPLAIN the Technical Specification requirements for placing an RPS trip unit in the tripped or bypassed condition within one hour.

EXPLANATION:

B/PT-102 provides an input to both 86A/PPLS and 86B/PPLS so both must be bypassed within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

"B" is correct and "A" is incorrect. "C" and "D" are incorrect because of the time limit and the logic combination.

KA#: 012000 2.1.32 Bank Ref #: 07-12-25 077 LP# / Objective: 0712-25 01.18 Exam Level: SRO-2 Cognitive Level: HIGH Source: BANK

Reference:

TS 2.15 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 091 Following a loss of coolant accident with severe core damage and high containment hydrogen concentration. A General Emergency has been declared and a Protective Action Recommendation made to evacuate all sectors to two miles and sectors M, N and O to five miles.

The Site Director has approved a TSC recommendation to operate the hydrogen purge system to lower containment hydrogen concentration. A projection of offsite doses has been made for the hydrogen purge release which is expected to last for six hours. The wind speed and direction has not changed and is expected to remain steady.

Which one of the following would require notification of the states and counties.

A. The dose assessment has been revised and the Emergency Classification downgraded to an "ALERT."

B. Command and Control has been transferred from the Control Room to the EOF.

C. Issuance of Potassium Iodide has been authorized for onsite personnel.

D. Non-Emergency Response personnel have been evacuated to North Omaha Station Question # 91 Revision: 0 KA #: 028000 2.4.30 Tier 2 Group 2 Hydrogen Recombiner and Purge Control System Knowledge of events related to system operations/status that must be reported to internal organizations or external agencies such as the State, the NRC or the transmission system operator.

Importance: 2.7 / 4.1 CFR Number: 5.43(b)(4)

Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Fort Calhoun Objective:

The person who completes this topic will be able to IDENTIFY the applicable definitions and variables that go into the making of Protective Action Recommendations EXPLANATION: A change to the EALs requires notification of the States and Counties. "A" is correct The distractors do not require notification.

KA#: 028000 2.4.30 Bank Ref #: N/A LP# / Objective: 1070-05 01.00 Exam Level: SRO-4 Cognitive Level: HIGH Source: NEW

Reference:

EPIP-EOF-7 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 092 Given the following plant conditions:

x Refueling is in progress x A new fuel assembly is being inserted into the core x Count rates on 2 Wide Range NI channels have doubled What action should be taken?

A. Stop fuel movement and initiate emergency boration in accordance with AOP-03, "EMERGENCY BORATION."

B. Stop fuel movement and evacuate containment in accordance with AOP-08, "FUEL HANDLING INCIDENT."

C. Withdraw the fuel assembly from the core and initiate emergency boration in accordance with AOP-03, "EMERGENCY BORATION."

D. Withdraw the fuel assembly from the core and evacuate containment in accordance with AOP-08, "FUEL HANDLING INCIDENT."

Question # 92 Revision: 0 KA #: 034000 K1.04 Tier 2 Group 2 Fuel Handling Equipment System Knowledge of the physical connections and/or cause-effect relationships between the Fuel Handling System and the following systems: NIS Importance: 2.6 / 3.5 CFR Number: 5.43(b)(7)

Fuel handling facilities and procedures.

Fort Calhoun Objective:

Use the Emergency Boration AOP to mitigate the consequences of an uncontrollable or unexplained positive reactivity addition.

EXPLANATION:

OP-12 precaution states that if the count rate doubles on 2 or more channels, withdraw the fuel assembly being inserted and Enter AOP-03. "C" is correct. "A" is incorrect because the fuel assembly is not withdrawn. "B" and "D" are incorrect because the wrong AOP is entered.

KA#: 034000 K1.04 Bank Ref #: N/A LP# / Objective: 0717-03 01.00 Exam Level: SRO-7 Cognitive Level: LOW Source: NEW

Reference:

OP-12 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 093 Given the following plant conditions:

x The plant is operating at full power x There is an unusually high amount of debris in the water x Level in all intake cells is 981 feet and lowering x Pressure is lowering on PI-1913A, "CONDENSER CIRC WATER PRESSURE INLET" What actions should be taken at this time?

A. Enter AOP-10, "LOSS OF CIRCULATING WATER," and throttle condenser outlet valves. Implement AOP-01, ACTS OF NATURE,Section IV, "Low River Water Level."

B. Initiate a plant shutdown using OP-4, "LOAD CHANGE AND NORMAL POWER OPERATION." Prepare for the potential loss of Fire Protection using SO-G-103, "Fire Protection Operability Criteria and Surveillance Requirements."

C. Initiate a plant shutdown using AOP-05, "EMERGENCY PLANT SHUTDOWN."

Prepare for the potential loss of Raw Water using AOP-18, "LOSS OF RAW WATER."

D. Trip the Reactor and enter EOP-00, "STANDARD POST TRIP ACTIONS." After completing these actions, trip the Circulating Water Pumps and enter AOP-10, "LOSS OF CIRCULATING WATER."

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL Question # 93 Revision: 0 KA #: 075000 A2.01 Tier 2 Group 2 Circulating Water System Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of intake structure Importance: 3 / 3.2 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

Use the Loss of Circulating Water Procedure to mitigate the consequences of a loss of cooling to the condenser or a Circulating Water System rupture.

EXPLANATION:

Choice "A" is correct per AOP-10. The other choices are incorrect because a plant shutdown is not directed under these conditions although it would be at lower river levels. Distractors address other concerns if the river level gets too low.

KA#: 075000 A2.01 Bank Ref #: N/A LP# / Objective: 0717-10 01.00 Exam Level: SRO-5 Cognitive Level: HIGH Source: NEW

Reference:

AOP-10, AOP-01 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 094 The Shift Chemist has just reported that the analysis of samples taken from both steam generators indicates that Dose Equivalent Iodine is 0.14 uCi/gm.

What action is required by Technical Specification 2.20, "Steam Generator Coolant Radioactivity" and why is this action required?

A. The plant must be placed in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to ensure that the radiological consequences of a steam line break accident are within 10 CFR 50.67 limits.

B. The plant must be placed in hot shutdown within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> to ensure that the radiological consequences of a steam generator tube rupture accident are within 10 CFR 50.67 limits.

C. The plant must be placed in hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to ensure that the radiological consequences of a steam line break accident are within 10 CFR 50.67 limits.

D. The plant must be placed in hot shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> to ensure that the radiological consequences of a steam generator tube rupture accident are within 10 CFR 50.67 limits.

Question # 94 Revision: 0 KA #: 000000 2.1.34 Tier 3 Group 4 Generic Knowledges and Abilities Knowledge of primary and secondary plant chemistry limits.

Importance: 2.7 / 3.5 CFR Number: 5.43(b)(2)

Facility operating limitations in the technical specifications and their bases.

Fort Calhoun Objective:

DISCUSS the basis for the LCOs (SRO only).

EXPLANATION:

"A" is the required action and basis for TS 2.20. "B" is incorrect, wrong accident. "C" and "D" have the wrong time limit.

KA#: 000000 2.1.34 Bank Ref #: N/A LP# / Objective: 0762-08 06.01 Exam Level: SRO-2 Cognitive Level: LOW Source: NEW

Reference:

TS 2.20 BASIS Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 095 The following plant conditions exist:

• The plant is at 90% during an increase in power.

• A CEA in Control Group 4 is discovered to be at 50 inches.

• All other Group 4 CEAs are at 120 inches.

• Attempts to move the rod are unsuccessful.

• I&C investigates and reports that there is an electrical fault in the rod drive panel and it will take 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> to repair.

• The PRC has determined that the CEA is trippable.

Which one of the following actions is required as a result of these conditions?

A. The power level increase and operations may continue without restrictions.

B. Align the remainder of the CEA's in Group 4 to within 12 inches of the inoperable CEA and do not raise power level above 90% until the CEA is repaired.

C. Align the remainder of the CEA's in Group 4 to within 12 inches of the inoperable CEA and reduce power level to 60% until the CEA is repaired.

D. Verify shutdown margin and be in HOT SHUTDOWN within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

Question # 95 Revision: 0 KA #: 000000 2.1.43 Tier 3 Group 4 Generic Knowledges and Abilities Ability to use procedures to determine the effects on reactivity of plant changes, such as reactor coolant system temperature, secondary plant, fuel depletion, etc.

Importance: 4.1 / 4.3 CFR Number: 5.43(b)(6)

Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects of core reactivity.

Fort Calhoun Objective: Discuss the power dependent insertion limit EXPLANATION: Technical Specification 2.10.2 requires realignment of all CEAs in a group to within 12 inches maintaining the CEA insertion limits. The COLR PDIL limits the reactor to 60% power with group 4 CEAs at 50 inches. "C" is correct. "A and "B" are incorrect because power must be reduced to 60%. "D" is incorrect because a shutdown is not required.

KA#: 000000 2.1.43 Bank Ref #: 07-62-08 LP# / Objective: 0705-09 01.13 Exam Level: SRO-6 Cognitive Level: HIGH Source: NRC 02 EXAM

Reference:

TS 2.10.2 Handout: PDIL CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 096 Which one of the following procedure changes can be processed as a temporary procedure change?

A. A change to the initial conditions in an Operations Survellance Test (OP-ST) procedure .

B. A change in a system Operating Instruction (OI) to allow manual operation of an automatic controller.

C. A change to an Abnormal Operating Procedure (AOP) to clarify entry conditions.

D. A change to an Emergency Operating Procedure (EOP) to require an additional action be taken due to a failed indicator.

Question # 96 Revision:0 KA #: 000000 2.2.06 Tier 3 Group 4 Generic Knowledges and Abilities Knowledge of the process for making changes in procedures.

Importance: 3 / 3.6 CFR Number: 5.43(b)(3)

Facility licensee procedures required to obtain authority for design and operating changes in the facility.

Fort Calhoun Objective:

Temporary procedure change implementation EXPLANATION:

"B" is correct because it is a non-change of intent change to an OI. "A" changes the intent of the procedure by changing initial conditions. "C" and "D" are incorrect because a temporary procedure change can not be used for AOPs or EOPs.

KA#: 000000 2.2.06 Bank Ref #: N/A LP# / Objective: 0762-08 10.03 Exam Level: SRO-3 Cognitive Level: HIGH Source: NEW

Reference:

SO-G-30 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 097 Given the following plant conditions:

x A radiological event has occurred x The Emergency Plan has been entered According to EPIP-EOF-11, " DOSIMETRY RECORDS, EXPOSURE EXTENSIONS AND HABITABILITY," The person in the "Command and Control" position must approve any authorized radiation dose greater than ________ in a year.

A. 1 rem B. 5 rem C. 10 rem D. 25 rem Question # 97 Revision: 0 KA #: 000000 2.3.04 Tier 3 Group 4 Generic Knowledges and Abilities Knowledge of radiation exposure limits under normal and emergency conditions.

Importance: 3.2 / 3.7 CFR Number: 5.43(b)(4)

Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Fort Calhoun Objective:

LIST the OPPD administrative exposure limits.

EXPLANATION:

According to EPIP-EOF-11, The Comand and Control Position must approve radiation dose greater than 5 rem in a year. "B" is correct the others are wrong.

KA#: 000000 2.3.04 Bank Ref #: N/A LP# / Objective: 1924-03 03.02 Exam Level: SRO-4 Cognitive Level: LOW Source: NEW

Reference:

EPIP-EOF-11 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 098 Given the following plant conditions:

x The Reactor is in Mode 4, Cold Shutdown x All Circulating Water Pumps have been secured x Raw Water Pumps AC-10A and AC-10C are operating x Radiation Monitor, RM-055, is inoperable x A sample has been taken of the contents of Waste Monitor Tank, WD-22B x The release rate calculations have been verified by two qualified individuals x A release permit has been prepared and approved by Chemistry What action (if any) must be taken before the Shift Manager can authorize a release of Waste Monitor Tank, WD-22B?

A. No additional action is required.

B. An additional Raw Water Pump must be started.

C. A Circulating Water Pump must be started.

D. An additional WD-22B sample must be taken.

Question # 98 Revision: 0 KA #: 000000 2.3.11 Tier 3 Group 4 Generic Knowledges and Abilities Ability to control radiation releases.

Importance: 3.8 / 4.3 CFR Number: 5.43(b)(4)

Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Fort Calhoun Objective:

Monitor Tank Releases EXPLANATION:

The requirements for the release are 2 samples independently verified by 2 qualified individuals. One more sample needs to be taken. "D: is correct and "A" is incorrect. One Circulating Water Pump or 2 Raw Water Pumps are required to be operating. Choices "B" and "C" are incorrect.

KA#: 000000 2.3.11 Bank Ref #: N/A LP# / Objective: 1950-04 10.01B Exam Level: SRO-4 Cognitive Level: HIGH Source: NEW

Reference:

CH-ODCM-0001 Handout:

CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 099 The following instructions are included as a contingency action in EOP-20, MVA-IA "If instrument air pressure is less than 90 psig, then IMPLEMENT AOP-17, LOSS OF INSTRUMENT AIR" If this is the case, how do you use AOP-17 in conjunction with EOP-20?

A. Exit EOP-20 and enter AOP-17.

B. Exit EOP-20 and enter AOP-17. Reenter EOP-20 when you reach the AOP-17 exit conditions.

C. Complete the actions in EOP-20. Enter AOP-17 when you reach the EOP-20 exit conditions.

D. Perform AOP-17 actions in parallel with EOP-20.

Question # 99 Revision: 0 KA #: 000000 2.4.08 Tier 3 Group 4 Generic Knowledges and Abilities Knowledge of how abnormal operating procedures are used in conjunction with EOPs.

Importance: 3.8 / 4.5 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

GIVEN a set of plant conditions, DETERMINE if the Standard Post Trip Actions (SPTA's), the Optimal Recovery Guidelines or the Functional Recovery Guideline (FGR) should be used.

EXPLANATION:

According to OPD 4-09, IMPLEMENTmeans to perform in parallel with. Therefore, "D" is correct and all of the distractors are incorrect.

KA#: 000000 2.4.08 Bank Ref #: 07-18-10 061 LP# / Objective: 0718-10 01.06 Exam Level: SRO-5 Cognitive Level: LOW Source: NRC 04 EXAM

Reference:

OPD 4-09 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL

CONFIDENTIAL NRC EXAM MATERIAL QUESTION NUMBER: 100 Given the following plant conditions:

Redacted - SUNSI Question # 100 Revision:0 KA #: 000000 2.4.16 Tier 3 Group 4 Generic Knowledges and Abilities Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines.

Importance: 3.5 / 4.4 CFR Number: 5.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal,and emergency situations.

Fort Calhoun Objective:

STATE the conditions that OCAG-1 is designed to mitigate.

EXPLANATION:

Redacted - SUNSI KA#: 000000 2.4.16 Bank Ref #: N/A LP# / Objective: 1074-01 02.02 Exam Level: SRO-5 Cognitive Level: LOW Source: NEW

Reference:

OCAG-1 Handout: NONE CONFIDENTIAL NRC EXAM MATERIAL