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Draft Written Exam
	ML102160334
Person / Time
Site:	Indian Point
Issue date:	04/09/2010
From:	Ferrick J; Entergy Nuclear Northeast
To:	Caruso J; Operations Branch I
	Hansell S
Shared Package
ML092470061	List: IR 05000247/2010301; ML101950011; ML102160274; ML102160314; ML102160334; ML102170047; ML102510238; ML102640428; ML102640438; ML102640564; ML102640579; ML102640589; ML102640597;
References
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{{#Wiki_filter:Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # 000007K102 Knowledge of the operational implications of the following concepts as they apply to the reactor trip: - Shutdown margin Importance 3.4 Question # 1 Which of the following statements is correct regarding boration to required concentration following a reactor trip from Mode 1 conditions? (Assume Tavg is stable at 547°F) A. Boration does not have to be considered because the online boron concentration is greater than the required shutdown boron concentration. B. Boration to the required shutdown concentration may be delayed up to 8 hours if reactor power had been less than 50% for the 48 hours prior to the trip. C. Boration to the required shutdown concentration may be delayed up to 8 hours if reactor power had been greater than 50% for the 48 hours prior to the trip. D. Boration to the required shutdown concentration must be commenced without delay regardless or power level prior to the trip. Answer: C Explanation/Justification: A. Incorrect but plausible because the online boraon concentration often does exceed the required shutdown concentration. B. Incorrect but plausible because an operator may be confused as to which power level allows for delay. C. Correct. D. Incorrect but plausible because an operator may assume no allowance for Xenon is allow in boration requirements. Technical

References:

2-POP-3.2 Proposed References to be provided: _N~o.::....n:...:...e=____________________

Learning Objective: 12LP-ILO-EOPS01 5 Question Source: Bank# - - - - IPEC Bank Note changes or Modified Bank # - - - - attach parent New x Question History: Last NRC Exam: Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 5 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO SRO Tier # 1 Group # 1 KIA # 000025K301 Knowledge of the reasons for the following responses as they apply to the Loss of Residual Heat Removal System: - Shift to alternate flowpath Importance 3.1 Question # 2 Given the following:

The plant is at reduced inventory in preparation for vacuum fill of the RCS following a mid-cycle RCP seal replacement.
The RCS is vented with the Pressurizer manway removed.
An RCS leak occurred
RCS level is decreasing
RHR flow is oscillating
The operating RHR pump was subsequently secured.
The crew is performing actions per AOP-RHR-1, Loss of RHR.

Which of the following identifies the reason why HCV 638, 21 RHR HX and HCV 640,22 RHR HX are initially left open for this condition? A. to provide a gravity feed path from RWST B. in preparation for restarting the pump C. to ensure ~P does not prevent subsequent opening the valves D. because motors are de-energized at reduced inventory Answer: A Explanation/Justification: A. Correct. If charging pumps cannot maintain RCS inventory, or a more rapid increase in level is desired, the RWST is gravity drained to the RCS via RHR. These valves must be left open.

B. Incorrect. Plausible because aligning the RHR to the RWST would provide subcooled water to the suction and the capacity of the RHR pumps would increase RCS inventory more rapidly. The pumps are started with the valves closed. C. Incorrect. Plausible because HCV-638 and 640 are butterfly valves and a small differential pressure could prevent the valves from opening. D. Incorrect. Plausible because during RHR operation the suction valves (730 and 731) are de-energized open at reduced inventory to prevent loss of RHR if the valves close. Note: 638 and 640 are not de-energized during RHR operation. Technical

References:

2-AOP-RHR-1

                                          ~---------------------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-AOPRHR 3 Question Source: Bank # ------ IPEC Bank Note changes or Modified Bank # ----- attach parent New x Question History: Last 2 NRC Exam2 at IPEC: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 5

                                                     ------~~------------

55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier# 1 Group # 1 KIA # 0000152236 Equipment Control - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. Importance 3.1 Question # 3 Given:

Plant heatup is in progress in accordance with 2-POP-1 .1, Plant Heatup from Cold Shutdown Conditions.
Final preparations are in progress to enter MODE 3.
24 RCP is in operation.
MCC 28 is de-energized for post maintenance testing of the normal supply breaker.
The Reactor Trip and Bypass Breakers are tagged out for Power Cabinet fuse clip replacement.

Can the unit enter MODE 3? A. Mode 3 cannot be entered. Tech Specs requires four RCS Loops OPERABLE to enter Mode 3 regardless of the status of the status of the Rod Control System. B. Mode 3 cannot be entered. Tech Specs requires two RCS Loops OPERABLE and one in operation to enter Mode 3 when the Rod Control System is NOT capable of rod withdrawal. C. Mode 3 can be entered. Tech Specs requires one RCS Loop OPERABLE as long as the Rod Control System is not capable of rod withdrawal to enter Mode 3. D. Mode 3 can be entered. Tech Specs requires two loops (RHR and/or RCS) OPERABLE and one in operation to enter Mode 3 regardless of the status of the status of the Rod Control System.

Answer: B Expla nation!Justification: A. Incorrect. Plausible because Tech Specs does require two RCS loops in operation when rod control is capable of rod withdrawal. Only one RCS loop must be in operation and another loop OPERABLE to enter mode 3. B. Correct. With MCC 28 de"energized a second RCP cannot be started (power to bearing lift pumps). Based on these conditions only one loop is OPEABLE and entry into mode 3 is not allowed. C. Incorrect. Plausible because the statement is true, only one loop is required to be in operation; however, a second loop must OPERABLE. D. Incorrect. Plausible because the combination of 2 loops is the LCO for MODE 4. Technical

References:

Technical Specifications Proposed References to be provided: None Learning Objective: 12LP"ILO"POP005 - 3 Question Source: Bank# IPEC Bank Note changes or Modified Bank # _ _ _ _ attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 5 55.43 (b) 2 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KlA# 000026K302 Knowledge of the reasons for the following responses as they apply to the Loss of Component Cooling Water: - The automatic actions (alignments) within the CCWS resulting from the actuation of the ESFAS Importance 3.6 Question # 4 The plant is in a normal full power lineup. During I&C troubleshooting, a technician inadvertently depressed Train B Containment Spray manual actuation pushbutton. What effect will this have on the Component Cooling Water System? I AuxCCW CCWfrom CCWPumps Pumps 0B valves RHRHx A. No Change No Change 4 valves closed No Change All Valves Both Valves i B. All Running All Running Closed Open One Valve C. All Running All Running 4 valves closed Open All Valves D. No Change No Change Closed No Change Answer: A Explanation!Justification:

A single manual pushbutton for Containment Spray will actuate a single train of Spray (1 pump and half of the valves). It will also actuate a single train of 08 isolation. A manual containment spray actuation will not actuate a safety injection. An automatic spray actuation signal will actuate safety injection if not already actuated. A. Correct. SI is not actuated and 4 of the 7 08 valves will close (those controlled by Train 8 08).

8. Incorrect. Plausible because the candidate may believe that a single manual pushbutton will actuate both trains of 08 valves and initiate an automatic SI.

C. Incorrect. Plausible because the candidate may believe that a single manual pushbutton will actuate one trains of 08 valves and initiate an one train of SI. D. Incorrect. Plausible because the candidate may believe that the manual pushbutton will actuate both trains of 08 valves. Technical

References:

Logic Diag 225105-1 Proposed References to be provided: None Learning Objective: 12LP-ILO-ESS001 - 5 Question Source: 8ank# - - - - IPEC 8ank Note changes or Modified 8ank # _ _ _ _ attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _-->('-...Ib)'--7_ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # 000027A215 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions:

                                                   - Actions to be taken if PZR pressure instrument fails high Importance          3.7 Question #      5 Given the following:

Unit is operating at 100% power.
A failure of the controlling pressurizer pressure channel caused actual pressurizer pressure to rise approximately 30 psig above normal.
Pressurizer Pressure Master Controller was placed in MANUAL.

Which ONE of the following describes the action required to reduce RCS pressure? A. Decrease the controller output. B. Increase the controller output. C. Lower the pressure setpoint adjustment. D. Raise the pressure setpoint adjustment. Answer: B ExplanationlJu stification: A. Incorrect and plausible an operator may think he has to lower the output to lower pressure B. Correct C. Incorrect but plausible because lowering the pressure setpoint would work in auto

D. Incorrect but plausible because of confusion between how the controller works in auto vs. manual Technical

References:

NA

                                        ~~-------------------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-RCSPZR - 9 Question Source: Bank# X IPEC Bank4233 Note changes or Modified Bank # _ _ _ _ attach parent New Question History: Last NRC Exam: NA

                                                        ~..,.:---

Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # 000029K206 Knowledge of the interrelations between the ATWS and the following: - Breakers, relays, and disconnects Importance 2.9 Question # 6 Reactor Trip and Bypass Breakers have been aligned to support testing of Reactor Protection Train A when an inadvertent Safety Injection signal is generated on Safety Injection Train A. Which of the following describes Reactor Protection System response? A. An ATWS occurs because Reactor Trip Breaker A and Reactor Trip Bypass Breaker B remain closed. B. An ATWS occurs because Reactor Trip Breaker B and Reactor Trip Bypass Breaker A remain closed. C. The Reactor trips. Reactor Trip Breaker A, Reactor Trip Breaker B, and Reactor Trip Bypass Breaker A are open. Reactor Trip Bypass Breaker B is racked out. D. The Reactor trips. Reactor Trip Breaker A, Reactor Trip Breaker B, and Reactor Trip Bypass Breaker B are open. Reactor Trip Bypass Breaker A is racked out. Answer: C Explanation/Justification: Meets KA 000029EK2.06 because the KA calls for knowledge of interrelations between trip breakers and ATWS. Since the question tests the knowledge of whether or not this breaker configuration can lead to an ATWS, the KA is met. A. Incorrect but plausible because an operator may think that the SI Train A only causes a Reactor Trip on RPS Train A

B. Incorrect but plausible because an operator may make the same error as for A plus not recall how trip and bypass breakers are paired. C. Correct D. Incorrect but plausible if an operator does not recall how trip and bypass breakers are paired. Technical

References:

System Description 16 Proposed References to be provided: None Learning Objective: 12Lp*ILO*!CROD

9 Question Source: Bank# !PEC Bank

                                               ----          Note changes or Modified Bank #     _ _ _ _ attach parent New                      x Question History:                     Last NRC Exam:       NA Memory or Fundamental Question Cognitive Level:             Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _--->(--/b)'--7_ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KJA# 000038A201 Ability to determine and interpret the following as they apply to a SGTR: - When to isolate one or more S/Gs Importance 4.1 Question # 7 Given:

Unit 2 has just experienced a steam generator tube rupture in 22 steam generator (SG).
E-O, Reactor Trip or Safety Injection, actions have led to a transition to E 3, Steam Generator Tube Rupture.
The operator has attempted to close 22 main steam isolation valve (MSIV). 22 MSIV failed to close.

What is the next action that must be taken to limit the release of radioactivity from 22SG? A. Continue attempts to close the 22 MSIV and cool down with all intact Steam Generators using Steam Generator Atmospheric Valves leaving remaining MSIVs open. B. Continue attempts to close the 22 MSIV and transition to ECA-3.1, SGTR with Loss of Reactor Coolant - Subcooled Recovery Desired. C. Isolate the remaining Steam Generators by closing their MSIVs and cool down using intact Steam Generator Atmospheric Valves. D. Isolate the remaining Steam Generators by closing their MSIVs and transition to ECA-3.1, SGTR with Loss of Reactor Coolant - Subcooled Recovery Desired. Answer: C Explanation/Justification:

A. Incorrect. Plausible because while not what is specified in E~3, the non return check valves will actually make this work. B. Incorrect Plausible because transition to ECA-3.1 would mitigate the event, but it is not what is specified in the EOP network. C. Correct. This method isolates remaining SG from ruptured SG. D. is plausible because transition to ECA-3.1 would mitigate the event, but it is not what is specified in the EOP network. Technical

References:

2-E-3 Proposed References to be provided: None Learning Objective: 12LP-ILO-AOPSG 1 ~ 2 Question Source: Bank#

                                                --- x- -   IPEC Bank 8668 Note changes or Modified Bank #      -------    attach   parent New Question History:                     Last NRC Exam:       NA Memory or Fundamental Question Cognitive Level:             Knowledge:                             X Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # 000054A102 Ability to operate and/or monitbr the following as they apply to the Loss of Main Feedwater (MFW): - Manual startup of electric and steam driven AFW pumps Importance 4.4 Question # 8 Given the following plant conditions:

Plant is at 50% power
21 Rod Drive MG set is out of service for bearing replacement
A fault occurs on 3A 480V bus, coincident with a unit trip Which ONE of the following statements describes the configuration of the auxiliary feedwater system, prior to any operator actions?

A. Only 22 and 23 AFW pumps running. AFW flow to 23 and 24 S/G's only B. All AFW pumps running, AFW flow to each S/G C. Only 21 and 22 AFW pumps running, AFW flow to 21 and 22 S/G's only D. Only 22 AFW pump running, NO AFW flow to any S/G Answer: A Explanation/Justification: A. Correct because shrink will cause 23 and 22 to start. 50% initial power makes it not obvious as to whether or not sufficient shrink will occur. However, it does occur in the simulator and the answers do not provide for any other possibility.

B. Incorrect but plausible. 21 AFW cannot run because of a fault on 3A. However it is plausible because there are safeguard loads on 22 EDG that have multiple breakers. Also it is easy to forget that the fault continues to keep the bus de-energized. C. Incorrect and plausible. 21 AFW running is plausible (see B.). 23 not running is plausible because of misunderstanding BO logic (note that U3 logic would support the 6A pump not running) D. Incorrect and plausible. This is plausible because of potential confusion with the BO logic and not knowing if sufficient shrink would have occurred. Technical

References:

System Description 27.1 Proposed References to be provided: _N_o_n_e_ _ _ _ _ _ _ _ _ _ __ Learning Objective: 12LP-ILO-MFW001 - 5 Question Source: Bank# X IPEC Bank - 8710 Note changes or Modified Bank # _ _ _ _ attach parent New Question History: Last NRC Exam: NA

                                                         -'--~--

Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _---->(.b)l-7 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # 000055A201 Ability to determine and interpret the following as they apply to a Station Blackout: Existing valve positioning on a loss of instrument air system Importance 3.4 Question # 9 A complete station blackout occurs. With no operator actions, how will the following plant equipment be affected? A. No Charging Pumps Running VC Monitors R-41/R-42 Supply/Return (PCV-1234, 1235, 1236, / 1237) - closed Diesel Generator Cooler Outlets (FCV-1176, 1176A) - open / B. No Charging Pumps Running Main Feedwater Regulating valves (FCV-417, 427, 437, 447) closed Bypass Feedwater Regulators (FCV-417L, 427L, 437L, 447L) open C. Atmospheric Dump Valves (PCV-1134-1137) - open CST to Hotwell Makeup (LCV-1128) - open Non-Regenerative Heat Exchanger (TCV-130) - open D. Pressurizer Spray valves (PCV-455A, 455B) - closed Loop Charging (204A1204B) - closed Charging Control (HCV-142) - closed Answer: A Explanation/Justification:

A. Correct B. Incorrect but plausible. An operator may misunderstand the failure position of the bypass FRVs C. Incorrect but plausible. An operator may misunderstand the failure positon of the ADVs. D. Incorrect but plausible. An operator may misunderstand the failure position of the loop charging isolations. Additionally, since trips/SI's without a complete station blackout will usually not lead to a loss of lA, an operator may pick a distractor for that reason as well. Technical

References:

System Description 27.1, 2-AOP-AIR-1 Proposed References to be provided: None Learning Objective: 12LP-ILO-EDS01 - 7 12LP-ILO-EDS01 - 17 Question Source: Bank# IPEC Bank Note changes or Modified Bank # _ _X_ _ attach parent4200 New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _----l.(..;....J.b):...-7_ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # 000056K104 Knowledge of the operational implications of the following concepts as they apply to Loss of Offsite Power: - Definition of saturation conditions. implication for the systems Importance 3.1 Question # 10 Given the following conditions:

The plant tripped due to a grid disturbance and loss of off-site power.
The crew is performing actions of ES-0.1, Reactor Trip Response.
RCS pressure is currently 2050 psig.
CETs indicate 628 degrees F and increasing slowly.

Which ONE of the following describes the conditions currently present in the RCS, and the status of natural circulation flow in accordance with ES-0.1, Reactor Trip Response, Attachment 3 Natural Circulation Verification? A. Saturated conditions; Natural Circulation flow in the RCS is established. B. Subcooled conditions; Natural Circulation flow in the RCS is established. C. Saturated conditions; Natural Circulation flow in the RCS is NOT established. D. Subcooled conditions; Natural Circulation flow in the RCS is NOT established. Answer: D Explanation/Justification: A. Incorrect. Not at saturation B. Incorrect. Natural Circ is not verified because subcooling is too low. Must be 19 degrees F C. Incorrect. Conditions do not indicate saturation

D. Correct. Saturation for 2050 psig is approximately 641 degrees F. Subcooling at this point is still 13 degrees but must be 19 degrees for Natural Circ verification Technical

References:

2-ES-0.2 Proposed References to be provided: ------------------------ ES-0.1 Attachment 3

                                                             ~-------

Learning Objective: 12LP-ILO-EOPEOO - 1 Callaway Question Source: Bank# 2007 IPEC Bank Note changes or attach parent Modified Bank #

                                                -----X      Attached New Callaway Last NRC Exam:

Question History: -2007 Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # 000057A104 Ability to operate and/or monitor the following as they apply to the Loss of Vital AC Instrument Bus: - RWST and VCT valves Importance 3.5 Question # 11 The following plant conditions exist:

Instrument Bus 21/21A has been lost due to inverter failure.
Pressurizer Level Channel 1 is in defeat.
Makeup Mode Selector switch is in AUTO What is the impact in the CVCS system due to the loss of Instrument Bus 21/21A?

A. Letdown isolation will occur. Automatic makeup will not occur. B. Valve 112B will open and 112C will close. Automatic makeup will not occur C. Letdown isolation will not occur. Automatic makeup will not occur. D. Valve 112B will open and 112C will close. Automatic makeup will occur. Answer: 0 ExplanationlJustification: A. Incorrect. Plausible because letdown isolation will not occur if PRZR Level Channel 1 is defeated B. Plausible because Charging pump suction will shift to RWST, but Auto makeup will occur. C. Incorrect. Plausible because Letdown Isolation will not occur, but Auto makeup will occur D. Correct

Technical

References:

2-AOP-IB-1 Attachment 1 Proposed References to be provided: None Learning Objective: 12LP-ILO-AOPIB1 - 1 Question Source: Bank# IPEC Bank Note changes or Modified Bank # _ _X_ _ attach parent 16763 New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _->{....... b)'--7_ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # 0000582236 Equipment Control - Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations. Importance 3.1 Question # 12 The following conditions exist at Unit 2:

The unit is in MODE 2 preparing for a Reactor startup.
Maintenance is performing troubleshooting on 21 Battery Charger due to log reading trends on charger output voltage.
21 Battery Charger trips and 21 DC Voltage is 108V.

Which one of the following actions is required? A. Shut down 21 Static Inverter. B. Transfer 21 Static Inverter to its Alternate Feed. C. Cross-connect 21 and 22 DC Buses. D. Open all reactor trip and reactor trip bypass breakers. Answer: B Explanation!Justification: A. Incorrect but plausible. 21 Static Inverter is shutdown if unable to transfer to alternate feed B. Correct C. Incorrect but plausible. Cross connecting DC busses is allowed only in Mode 5. D. Incorrect but plausible. An operator may believe this is necessary since the unit is in Mode 2. Technical

References:

2-AOP-DC-1

Proposed References to be provided: None Learning Objective: 12LP-ILO-AOPDC1 -1 Question Source: Bank# X IPEC Bank 16628 Note changes or Modified Bank # _ _ _ _ attach parent New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # 0000622242 Equipment Control- Ability to recognize system parameters that are entry level conditions for Technical Specifications. Importance 3.9 Question # 13 Which of the following events are entry conditions for Technical Specifications?

1. One Non-Essential Service Water Pump inoperable

2. One Essential Service Water Pump inoperable

3. TCV-1103 Containment Building Air Temp Control Valve fails closed

4. FCV-1176 Diesel Generator Cooling Water fails closed

5. NPO finds FCV 1111, SWP 24/25/26 SUP TO CONV NON ESSEN STOP and FCV 1112, SWP 21/22/23 SUP TO CONV NON ESSEN STOP open 50% each

6. Swapping from 1, 2, 3 header to 4, 5, 6 header as essential A.

2,4,5 (5,) w . B. 1,2,6 C. 2,4,6 D. 2,3,4 Answer: A ExplanationlJustification:

1. Incorrect. TS requires 2 NESW pumps

2. Correct

3. Incorrect. TCV-1103 is not required (TCV-1104 and 1105 are required)

4. Correct

5. Correct Essential and Non-Essential headers are required to be isolated from each other

6. Incorrect. Note allows 8 hours to split headers.

A. Correct

B. Incorrect C. Incorrect D. Incorrect Technical

References:

Tech Spec 3.7.8 Proposed References to be provided: None Learning Objective: 12LP-ILO-SW001 - 10 Question Source: Bank # IPEC Bank

                                           ----        Note changes or Modified Bank #   _ _ _ _ attach parent New                   x Question History:                 Last NRC Exam:      NA Memory or Fundamental Question Cognitive Level:         Knowledge:                            X Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 (b) 5 55.43 (b) 2 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # 000077AK 103 Knowledge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances: - Under excitation Importance 3.3 Question # 14 Given the following conditions:

Plant is at 100% power 1070 MWe
Generex is in AC Control
Generator H2 Pressure is 60 psig The System Operator has notified the plant that system grid voltage is high and forecasted to go higher.

If the System Operator requests the plant to take in the maximum value of MVARs to help stabilize the grid. Using Graph EL-1, what is the maximum allowed MVAR incoming value, and how is the adjustment made? MAX INCOMING VALUE METHOD OF ADJUSTMENT I I A. 410 MVARs AC Raise/Lower Switch B. 410 MVARs DC Raise/Lower Switch I C. 490 MVARs AC Raise/Lower Switch D. 490 MVARs DC Raise/Lower Switch Answer: A Explanation/Justification:

A. Correct. Candidate must use the limit of the Under Excited Reactive Ampere Limit (URAL) to determine Maximum VARs IN versus the generator hydrogen pressure. B. Incorrect. Plausible because the reactive load value is correct; however, the method of adjustment is incorrect with the GENEREX in AC Control. Adjustments using the DC Raise Lower Switch will be corrected to AC setpoint when in AC control. C. Incorrect. Plausible because the reactive load value is incorrect but it is the value obtained if the candidate uses the hydrogen pressure curve instead of the URAL curve; the method of adjustment is correct with the GENEREX in AC Control. D. Incorrect. Plausible because the reactive load value is incorrect but it is the value obtained if the candidate uses the hydrogen pressure curve instead of the URAL curve; the method of adjustment is incorrect with the GENEREX in AC Control. Technical

References:

Graph EL-1 Proposed References to be provided: Graph EL-1 Learning Objective: 12LP-ILO-MTG02 - 8 12Lp*ILO-MTG02 - 2 Question Source: Bank# ---- IPEC Bank Note changes or Modified Bank # _ _X_ _ attach parent New Watts Question History: Last NRC Exam: Bar 2008 Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: x 10 CFR Part 55 Content: 55.41 55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier# 1 Group # 1 KIA # OOWE04K302 Knowledge of the reasons for the following responses as they apply to the LOCA Outside Containment: Normal, abnormal and emergency operating procedures associated with LOCA Outside Containment Importance 3.4 Question # 15 Given the following:

A LOCA outside containment has occurred .
The crew is performing the actions in ECA-1.2, LOCA Outside Containment.

Which ONE (1) of the following actions will be attempted to isolate the break and which indication is used to determine if the leak has been isolated in accordance with ECA-1.2? A. Isolate SI Hot Leg Injection piping; PZR level is monitored, because with the break isolated, RCS inventory will rapidly rise. B. Isolate SI Hot Leg Injection piping; RCS pressure is monitored, because SI flow will repressurize the RCS with the break isolated. C. Isolate RHR Cold Leg Injection piping; PZR level is monitored, because with the break isolated, RCS inventory will rapidly rise. D. Isolate RHR Cold Leg Injection piping; RCS pressure is monitored, because SI flow will repressurize the RCS with the break isolated. Answer: D Expla nation/Justification:

A. Incorrect. Plausible because candidate must remember that all Hot Leg Injection piping is inside VC. Also Inventory is not the condition monitored to determine if leak isolation is successful. B. Incorrect. Plausible because candidate must remember that all Hot Leg Injection piping is inside VC. Also RCS pressure is the parameter monitored to determine if leak isolation was successful. C. Incorrect. Plausible because RHR Cold Leg Injection piping is located outside the VC; however, Inventory is not the condition monitored to determine if leak isolation is successful. D. Correct. Some RHR Cold Leg Injection piping is located outside the VC and RCS pressure is the parameter monitored to determine if leak isolation was successful. Technical

References:

2-ECA-1.2 BG

                                            ~~~~~-----------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPC12 - 3 Question Source: Bank# _--..:.....:X__ IPEC Bank 16661 Note changes or Modified Bank # _ _ _ _ attach parent New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KJA# 00WE05K201 Knowledge of the interrelations between the Loss of Secondary Heat Sink and the following: - Components, and functions of control and safety systems, including instrumentation, ~als, inut.....'ffO'Cl<S;failure modes, and automatic and manual features Importance 3.7 Question # 16 Following a small-break LOCA and SI actuation, core exit TC's read 625°F and increasing.

   * . RCS pressure is 1400 psia and rising.

5/G pressures are stable at 900 psig.
Containment pressure is stable at 3 psig.
The control room operators are attempting to establish MBFW flow in response to a loss of secondary heat sink.
They are unable to lift the live lead on the feed water isolation relay signal.

Which one of the following describes the plant response? A. The MBFW pumps cannot be reset to provide flow. B. The 51 signal cannot be reset. C. The MBFW regulator valves cannot be opened from the control room. D. Establishing MBFW flow will result in an excessively rapid RCS cooldown and depressurization. Answer: C ExplanationlJustification:

A. Incorrect and plausible since an operator may confuse the function of lifting the relay lead belieiving it will affect MBFPs. B. Incorrect but plausible since and operator may confuse function of lifting lead believing it affects SI reset. C. Correct - with this lead in place, FRVs will not open due to the Reactor Trip with Low Tavg signal. D. Incorrect but plausible because an operator may believe that failure to lift the lead could cause FRVs to be full open. Technical

References:

2-FR-H.1 Proposed References to be provided: Learning Objective: 12LP-ILO-EOPFH 1 - 5 Question Source: Bank# x

                                                   --------   IPEC Bank 8170 Note changes or Modified Bank #        --------   attach  parent New Question History:                     Last NRC Exam:          NA Memory or Fundamental Question Cognitive Level:             Knowledge:
                                     ~~~~:s~enSion or                     @r~

10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 1 KIA # OOWE11A101 Ability to operate and/or monitor the following as they apply to the Loss of Emergency Coolant Recirculation: - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Importance 3.9 Question # 17 Given:

A LBLOCA has occurred.
Operators are performing ECA-1.1, Loss of Emergency Coolant Recirculation.
Containment pressure is 24.3 psig and is lowering slowly.

Which of the following describes how the Containment Spray system will be operated, and why? The Containment Spray System is operated as directed in ... A. ECA-1.1 beoousQ it establisbes minimum required eOFltaifilTlent spray flow andconseruesRWSTinwnter;'. flc(q..;),f ,,,,,,, ,.,I"I,}J) t~W'.}"'P..a

                   ....                  ...... !J.,..: )   III' ;),-     ft c (jJ V-,,/It':    ? "l. ~ l. ( i N"{- /t.. 7' B. FR-Z.1, Response to High Containment Pressure, since restoration of the critical safety function takes precedence.

C. ECA-1.1 since FRPs (Functional Restoration Procedures) are NOT implemented during the performance of ECA-1.1. D. FR-Z.1 because oentaiflfflent is the last fission prQduot barrier aetions ale the same as S.A,MGs (Severe Accident Management.Guidelines) Et 1-1 / (l ;') l. ) r./o,- 4 J1prt'; U (c! ~r JPIl-11.. Answer: -A- -

Explanation/JusUfication: A. Correct B. Incorrect. Plausible because FR procedure typically have a higher priority than other emergency procedures. There are two procedures (ES-1.3 Transfer to Cold Leg Recirculation and ECA-O.O Loss of All AC Power) that take priority over FRPs. The concept that a procedure may take prioity over the FRPs is not unrealistic. ECA-1.1 addresses a "best guess" approach to containment conditions where FR-Z.1 addresses a "worst case" approach. The need to conserver RWST inventory for core cooling is a higher priority C. Incorrect. Plausible becauseECA-1.1 is a special case and candidates may believe that no FRPs are implemented. FRP are implemented in ECA-1.1 if the condition occurs with the exception of FR-Z.1 due to the need to conserve inventory for core cooling. D. Incorrect. Plausible because containment is the final fission product barrier. In general actions in the SAMGs are focused on maintaining containment intact. Technical

References:

2-FR-Z.1 Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPC12 - 6 Question Source: Bank# X

                                              --------    IPEC Bank 24215 Note changes or Modified Bank #                attach parent New Question History:                    Last NRC Exam:       NA Memory or Fundamental Question Cognitive Level:            Knowledge:                              X Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 _ _~(b::...L)-=.5_ _ _ __ 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier# 1 Group # 1 KIA # 00WE12K202 Knowledge of the interrelations between the Uncontrolled Depressurization of all Steam Generators and the following: Facility's heetJ:a.wgya1 sy.s1ems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Importance 3.6 Question # 18 During the performance of ECA-2.1, Uncontrolled Depressurization of All Steam Generators, the following plant condition exists:

Cooldown rate of the RCS is greater than 100°F/hr How is the crew directed to control feedwater flow?

A. Feedwater flow is terminated to all but a single intact SG, which is fed at 85gpm B. Feedwater flow is reduced to 85 gpm to each SG with narrow range level less than 9%. C. Feedwater flow is preferentially maximized to 22 or 23 SG until narrow range is > 10% D. Total feedwater flow is maintained at 400 gpm until narrow range level in any SG is > 10% Answer: B Explanation!Justification:

A. Incorrect. Plausible because a similar strategy is used in FR-H.1 for

feeding a hot dry SG. B. Correct C. Incorrect. Plausible because a similar strategy is used in several emergency procedures to ensure steam for the turbine driven AFW pump. D. Incorrect. Plausible because 400 gpm is the minimum normal feedwater

flow rate to ensure adequate heat sink if level is < 10%.

Technical

References:

2-ECA-2.1 Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPC21 - 4 Question Source: Bank# - - - - IPEC Bank Note changes or Modified Bank # _ _ _ _ attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 2 KIA # 000001A201 Ability to determine and interpret the following as they apply to the Continuous Rod Withdrawal: - Reactor tripped breaker indicator Importance 4.2 Question # 19 Unit 2 is operating at 100% power

Control Bank D rods start stepping out due to a Logic Cabinet Malfunction
Operators manually trip the Reactor.
90 seconds after the initial pressing of the Reactor Trip Push Button, the button is pressed a second time during the read-through of E-O, Reactor Trip or Safety Injection.

After this action the following indications are observed:

RTA - Green Light Lit
RTB - No Lights Lit (bulbs and sockets are working correctly)

Based on these indications, which of the following is the next appropriate action to be taken by the team? A. Manually insert control rods. B. Dispatch NPO to locally trip the Reactor C. Initiate Emergency Boration of the RCS D. Verify Turbine Trip Answer: D

(irA? Explanation/Justification: " This question requires the andidate to realize that 1)RTB did not open, but the reactor is tripped by RT 2) That an automatic trip signal exists against RTB from SG low low level 90 seconds after trip from 100% power, and 3) That with this is the proper breaker indication for RTB in these circumstances. A. Incorrect but plausible. Plausible because if the reactor was not tripped the next action would be to insert control rods per FR-S 1 . B. Incorrect but plausible. Plausible because a candidate may believe this is done even if the reactor did trip or per FR-S1 if it did not. C. Incorrect but plausible. Plausible because an operator may believe the reactor is not tripped and this action would be taken in FR-S1. Also plausible because an operator may believe this will be done in ES-0.1 due to breaker indications with a tripped reactor. D. Correct. The reactor is tripped with one breaker open. The next procedure step is to verify turbine trip. Technical

References:

System Description 16.1 Proposed References to be provided: None Learning Objective: 12LP-ILO-ICROD - 10 Question Source: Bank# ---- IPEC Bank Note changes or Modified Bank # attach pa rent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _------>(--<b):....-6_ _ _ __ 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 2 KIA # 000032A203 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: - Expected values of source range indication when high voltage is automatically removed Importance 2.8 Question # 20 Unit 2 is in MODE 3 with the following conditions:

Tave 54rF
Pressurizer pressure 2235 psig
Reactor trip breakers CLOSED
Source range counts 52 cps (N31) and 55 cps (N32)
Source range HV Manual On/Off is in NORMAL
ALL Control Rod Banks are INSERTED
ALL Shutdown Rod Banks are Withdrawn to 223 Steps An I&C technician is troubleshooting power source problems with the NIS drawers that were noted a few days earlier following a reactor trip. During the troubleshooting activities, the following indications are received at the main control boards:
SOURCE RANGE LOSS OF DETECTOR VOLT actuates.
Source range counts: 52 cps (N31), 0 cps (N32)
Reactor trip breakers CLOSED Which ONE of the following describes what the I&C technician did?

A. Removed the CONTROL POWER fuses for N32 with the Level Trip switch in BYPASS B. Removed the INSTRUMENT POWER fuses for N32 with the Level Trip switch in BYPASS.

C. Activated the RPS input for the SOURCE RANGE BLOCK. D. Removed power simultaneously to TWO Power Range channels. Answer: B Explanation/Justification: A. Incorrect. Removal of CONTROL POWER fuses will not result in loss of high voltage to the detector (supplied from the Instrument Power fuses) and thus would not result in loss of N32 indication. Regardless of Level Trip Bypass switch position, removal of the control power fuses would result in reactor trip from this condition. B. Correct. INSTRUMENT POWER supplies the High Voltage for the detector. Control power supplies power to the protection bistables including the Level Trip Bypass circuit. C. Incorrect. Plausible because RPS input for SR Block de-energizes High Voltage to both SR NIS channels. Indication that only one channel has lost indication should eliminate this distractor. Candidate must distinguish between conditions that result in de-energization of a single channel and both channels. D. Incorrect. Plausible because removal of power to 2 PR NIS channels will result in removal of High Voltage to both SR NIS. This can be bypassed using the SR HV Manual On/Off switches. Candidate must recognize that the HV Manual On/Off switch is not in the correct position to allow de energizing more than one power range instruments Technical

References:

System Description 13.1 Proposed References to be provided: None Learning Objective: 12LP-ILO-ICEXC - 7 12LP-ILO-ICEXC - 9 Question Source: Bank# - - - - IPEC Bank Note changes or Modified Bank # - -X- - attach parent New DC Cook Question History: Last NRC Exam: 2002 Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X

10 CFR Part 55 Content: 55.41 (b) 7 55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 2 KIA # 000033A103 Ability to operate and!or monitor the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: - Manual restoration of power Importance 3 Question # 21 Given the plant is at 100% power. Intermediate Range Channel N-35 was removed from service due to a failure of the high voltage power supply. I&C has completed repairs to Intermediate Range Channel N-35. The major actions to return N-35 to service are listed below.

1. Install Instrument and Control Power Fuses (warm up for 30 minutes)

2. Verify Level Trip Switch is in BYPASS

3. Remove blocking strips for Reactor Trip and Rod Stop

4. Perform Bistable setpoint verification.

5. Place Level Trip Switch to NORMAL Which of the following identifies the proper sequence for restoration of Intermediate Range Channel N-35?

A. 1, 2, 3, 4, 5 B. 2,1,5,3,4 C. 1, 3, 2, 5, 4 D. 2,1,4,3,5 Answer: 0 Explanation!Justification:

Verify Level Trip Switch is in BYPASS. Should be first action to prevent possible trip Install Instrument and Control Power Fuses (warm up for 30 minutes). Second to energize the drawer Perform Post Maintenance Testing. After drawer is energized then ensure it functions properly Remove blocking strips for Reactor Trip and Rod Stop. After drawer is verified to be functioning properly then remove blocking strips. Place Level Trip Bypass Switch to Normal. The final step to return to service A. Incorrect B. Incorrect C. Incorrect D. Incorrect Technical

References:

2-S0P-13.1 ~~ None'

                                                                       'f Proposed References to be provided:

Learning Objective: 12LP-ILO-ICEXC -11 Question Source: Bank# IPEC Bank Note changes or Modified Bank # ---- attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 2 KIA # 000036K101 Knowledge of the operational implications of the following concepts as they apply to Fuel Handling Incidents: - Radiation exposure hazards Importance 3.5 Question # 22 Given the following conditions:

Irradiated fuel assemblies are being shuffled in the Spent Fuel Pool (SFP),
An irradiated fuel assembly has been lifted clear of the racks and is in transit toward its new assigned position,
SFP level is noted to be dropping slowly,
The Fuel Transfer Canal Gate is closed and latched.

Which ONE of the following descr1bes the preferred course of a<f.\on regarding-,t' . -..../ the irradiated fuel assembly1A-tAJ 6t Aof..;-ptt- I) M ~ PI- rt eYH fJ.' -t v~ f 5 ~[ A. 8. Place the assembly in an appropriate location.

Return the assembly to its original location in the rack .

~~

C. Lower the assembly to the bottom of the SFP and check the gate seal inflated. D. Continue moving the assembly toward the new location and check the gate seal inflated. Answer: A ExplanationlJustification: A. Correct. AOP-FH-1 directs this action

8. Incorrect. Plausible because candidates may be concerned with SFP Zone requirements.

C. Incorrect. Plausible because this action is similar to an action used inside the VC. D. Incorrect. Plausible because candidates may be concerned with SFP Zone requirements.

Technical

References:

2-AOP-FH-1 Proposed References to be provided: None Learning Objective: 12LP-ILO-FHD001 - 11 Question Source: Bank# - -x- - IPEC Bank 23994 Note changes or Modified Bank # ---- attach parent New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _--->(--I.b):...-7_ _ _ __ 55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 2 KIA # 0000032431 Emergency Procedures/Plan Knowledge of annunciators alarms, indications, or response instructions. Importance 4.2 Question # 23 The following conditions exist on Unit 2:

Power is at 75% during a power ascension.
Rods were being withdrawn to maintain Tavg on program.
The Rod Drive shaft disconnected from Control Bank C Rod C3.
The RCCA has fully inserted into the fuel assembly guide tubes.

Which of the following identifies the Alarms expected for this event? I NIS Power I C3 Rod Bottom Rod Bottom Range Dropped Rod Control Light Rod Stop Rod Rod Stop Urgent Failure A. ON ON ON OFF B. OFF ON OFF ON \ C. ON OFF OFF ON [ \ D. OFF OFF ON OFF Answer: D Explanation/Justification:

C3 Rod Bottom Light comes from IRPI which actually measures drive shaft position. Since the drive shaft is still fully withdrawn, this light will be OFF Rod Bottom Rod Stop comes from IRPI which actually measures drive shaft position. Since the drive shaft is still fully withdrawn, this alarm will be OFF NIS Dropped Rod Rod Stop is generated from Power Range NIS decreasin'g at greater than 5% in 5seconds. This alarm should be ON Rod Control Urgent Failure is not expected when the rod drops; however, this alarm will be on for most dropped rod recovery. A. Incorrect B. Incorrect C. Incorrect D. Correct Technical

References:

Proposed References to be provided: Learning Objective: 12LP-ILO-AOPROD - 6 Question Source: Bank# IPEC Bank Note changes or Modified Bank #

                                                  - -x- -        attach parent New Braidwood Question History:                     Last NRC Exam:           2002 Memory or Fundamental Question Cognitive Level:             Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _---->.(-t'b)_2_ _ _ __ 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 2 KIA # 0000742225 Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. Importance 3.2 Question # 24 Which of the following describes a basis for LCO 3.5.2, 'ECCS OPERATING'? A. Three of the four accumulators provide enough water to fully recover the core before significant clad melt following a LOCA. B. The boron concentration in the RWST prevents a return to criticality event following a main steam line break. C. Maximum hydrogen generation from zirconium water reaction is ::::;0.17 times the hypothetical amount generated if all zirconium were to react. D. Three ECCS trains are required to ensure that sufficient flow is available, assuming a single failure affecting anyone train. Answer: D Explanation!Justification: A. Incorrect. Plausible because three of four accumulator will PARTIALLY recover the core before significant clad melt following a LOCA. B. Incorrect. Plausible because the candidate must recognize that RWST is a separate LCO and the boron concentration limits the potential for a post trip return to criticality and achieve significant power. . C. ncorrect. Plausible because the maximum hydrogen generation is part of the basis of this LCO; however the limit is 0.01 times the hypothetical amount D. Correct Technical

References:

Technical Specifications 3.5.2 Basis Proposed References to be provided: None Learning Objective: 12LP-ILO-SIS01 - 11

Question Source: Bank# IPEC Bank Note changes or Modified Bank # attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 2 KIA # OOWE03K201 Knowledge of the interrelations between the LOCA Cooldown and Depressurization and the following: - Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features Importance 3.6 Question # 25 The following plant conditions exist while the team is responding to SI reduction to reduce RCS injection flow during the performance of ES-1.2, Post LOCA Cooldown and Depressurization, for a small break LOCA.

One charging pump is running
Both RHR pumps are secured

    * # 22 SI Pump is running
    * # 24 RCP is running

Containment pressure is 1.2 psig
RCS Hot Leg temperatures are 330°F and trending down
RCS subcooling is 110°F and trending up slowly
Pressurizer level is 38% and stable The team is evaluating conditions to stop the remaining SI pump.

Which of the actions below should the team take first at this time? A. Stop #22 SI Pump B. Start one RHR pump C. Depressurize the RCS to refill the Pressurizer D. Manually operate SI pumps as necessary Answer: B

Exp la nation!Justification: A. Incorrect. Plausible because subcooling is a significant value; however, the required subcooling for these conditions is 209. Candidate should recognize that inadequate subcooling exists. B. Correct. Starting an RHR pump in "injection mode" will allow securing 22 SIP without the required subcooling as long as hot leg temperature is < 345 degrees F. This temperature ensures saturation conditions in the RCS are below the shutoff head of RHR pumps. C. Incorrect. Plausible because this action would be performed if Pressurizer level was < 28%. Candidate should recognize that adequate pressurizer level exists. D. Incorrect. Plausible because this action is on the foldout page if conditions degrade and require SI reinitiation. Candidate should recognize that conditions are not degrading (Pressurizer level stable and subcooling trending up). Technical

References:

2-ES-1.2 Proposed References to be provided: Learning Objective: 12LP-ILO-EOPS12 - 1 Question Source: Bank #

                                              - -x- -        IPEC Bank 24344 Note changes or Modified Bank #   ------        attach  parent New Question History:                    Last 2 NRC Exams at IPEC:           NA Memory or Fundamental Question Cognitive Level:            Knowledge:

Comprehension or Analysis: x 10 CFR Part 55 Content: 55.41 (b) 10 55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 2 KIA # OOWE06K202 Knowledge of the interrelations between the Degraded Core Cooling and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility Importance 3.8 Question # 26 The following plant conditions exist:

Inadequate core cooling conditions exists
Operators unable to re-establish high pressure SI flow M
All Core Exit Thermocouples indicate greater than 1200°F ~*jnH'l

    /I A_*     Unable to establish RCP restart criteria           / '/.k'/~~rs   r'      '7:'
    ~~

e =J=.Se reeoilirtlends the operator~

                                     ~                           ~            -,

j.fo.t~j5.. start the RCPs (one at a time) until CE s If Indicate less than 1200°F. Is the TgCs recommendation to restart the RCPs 1# f correct for these plant conditions? A. NO, RCP start without adequate support conditions will result in seal failures and greater loss of inventory. r

                                                                                    ~ft B. NO, RCP start will result in phase separation causing a deeper un of the core.

C. YES, RCP start should be done regardless of support conditions since a seal failure LOCA would aid in event mitigation. D. YES, RCP start should be done regardless of support conditions to extend the time in wAioA core damag~1 occ~ be-Mit£-

              ~,1>~ ~.~ ISC .d~~;{, ~

iA.efCJP<; ~ it: ~ ~

                            ~.~~ . .s~

Answer: D Explanation/Justification: A. Incorrect but plausible. A candidate would believe we should not start RCPs because there are situations where we do not start RCPs, but that is based on SG level. B. Incorrect but plausible. Same explanation as A. C. Incorrect but plausible. LOCA flow often aids in heat removal, but not in this case. FR-C.1 conditions cannot be met without a loss of inventory, so this LOCA is unlikely to help. D. Correct per EOP background Technical

References:

2-FR-c.1 Background Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPC01 - 1 Question Source: Bank# - -x- - IPEC Bank - 8481 Note changes or Modified Bank # attach parent New

                                       'J...
                                  / Las~RC Exa~ '"N7!r Question History:

Question Cognitive Level: ( r Memory or Fundamental Knowledge: PO Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 5 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier# 1 Group # 2 KIA # 00WE08K301 Knowledge of the reasons for the following responses as they apply to the Pressurized Thermal Shock: - Facility operating characteristics during transient conditions, including coolant chemistry and the effects of temperature, pressure, and reactivity changes and operating limitations and reasons for these operating characteristics Importance 3.4 Question # 27 15 minutes ago, the plant experienced a main steamline break, from 100% power. Because of difficulties in closing the MSIVs, 23 and 24 SGs have blown dry. Current Plant status is as follows:

E-2, Faulted SG Isolation, is in progress.
RCS temperature is 290°F and decreasing.
SI flow is still being supplied to the RCS.
Total AFW flow is 800 gpm.
All RCPs have been stopped due to loss of cooling water.
RCS Pressure is 1000 psig and steady.
Attachment 1 of E-O, Reactor Trip or Safety Injection, has been completed.

Which of the following is of greatest immediate concern? A. A crack could propagate in the reactor vessel wall due to a pressurized thermal shock event. B. Injection of ECCS accumulator nitrogen into the RCS is imminent, natural circulation cooling will be limited.

C. Controlled cooldown will be a challenge when 24 RCP is started due to 23 and 24 MSIVs being open. D. The loss of thermal driving head in the dry SGs will reduce the amount of natural circulation flow, due to stagnant coolant loops. Answer: A Explanation/Justification: A. Correct. Red Path conditions have just been met for PTS. With this amount of AFW flow and SI flow, the plant is causing a big challenge in this safety function. B. Incorrect but plausible. Accumulator injection is imminent and N2 injection could hamper natural circulation flow, but this is not the greatest immediate concern. C. Incorrect but plausible. Eventually (after soak) 24 RCP should be started if available. If 23 and 24 MSIVs are not shu! tRi¥ will pose a challen~e, but it. --.1.A is not the greatest immediate concern. ~ rLt~ ~t{ /1£:1 ~ ~...,., D. Incorrect but plausible. This statement is correct, but cooling at this pOint t { in the event is not an issue since the main challenge is due to too much cooling. Technical

References:

FR-P.1 Status Tree

                                          ~--------------------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPFP1 - 3 Question Source: Bank# --~-- X IPEC Bank 24216 Note changes or Modified Bank # attach parent New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _-.l.(..::...Lb)~2:.-_ _ __ 55.43 (b)

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 003 Reactor Coolant Pump KIA # Knowledge of the physical connections and/or cause effect relationships between the RCP and the following 003 K1.12 systems CCW Importance 3.0 Question # 28 The plant is at 100% power. t...

                 /
                                   ---*()~

CCW flow i los 22 RC~
Upper beari emperature 176°F and rising at 5°F/minute.
Lower bearing temperature 186°F and rising at 5°F/minute.

Seal injection flow has been maintained to the RCP. Which ONE of the following describes the MAXIMUM time allowed before the crew must stop the 22 RCP? A. 1530 B. 1532 C. 1533 D. 1535 Answer B Explanation/Justification: A. Incorrect. Plausible because the bearing temperatures are close to but below the trip setpoints.

B. Correct because the procedure (2-AOP-CCW-001 step 4.3 of Rev. 3) specifies tripping the RCP if CCW is lost for 2 minutes. C. Incorrect. Plausible because at this point the lower bearing temperature will exceed 200F D. Incorrect Plausible because at this point both bearing temperatures will exceed 200F. Technical

References:

2-AOP-CCW-1 (pg 7) Proposed References to be provided: Learning Objective: 12LP-ILO-RCSRCP 10 Question Source: Bank# IPEC Bank 19162 Note changes or Modified Bank # X attach parent New Question History: Last NRC Exam: Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) (4) 55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # -0::-:0~4-::C-OOO-K-50-7 Knowledge of the operational implications of the following concepts as they apply to the CVCS: - Relationship between SUR and reactivity Importance 2.8 Question # 29 Given the following:

Unit 2 has run for two months after completing a refueling outage when a unit trip occurs.
24 hours later, a Reactor Startup is in progress.
Shutdown Banks have been withdrawn and some Rod Control system troubleshooting is in progress which is delaying pulling the Control Banks.
The ATC notices that SR SUR is negative and cOllnt rate is rapidly decreasing.

Which of the following is the cause? A. Swapping charging pumps to a pump that was in service 30 days ago. B. Excessive check valve leakage during a safety injection pump surveillance test. C. Placing a CVCS mixed bed demineralizer in service that was last used during the refueling outage. D. Swapping PZR backup heaters to a group that was last used during the refueling outage. Answer: C Explanation/Justification: A. Incorrect but plausible because an operator may not remember the boron letdown curve for BOL

B. Incorrect but plausible since the SI pumps circulate RWST water which has sufficient boron concentration to affect RCS temp. The choice is incorrect because at NOP, SI pumps should not affect RCS parameters. C. Correct reference AOP-UC-1 and system descriptions. D. Incorrect but plausible. Heaters may affect RCS temperature because of boron coming plating and being released from heater. Also changes in PZR citealatiorr rate can affect Tavg, however 2 degrees F is excessive for this change in configuration. A knowledgeabre license candidate should know this is excessive. Technical

References:

Proposed References to be provided: None Learning Objective: 12LP-ILO-CVCS - 15 Question Source: Bank#

                                              - - - - IPEC Bank Note changes or Modified Bank #                attach parent New                     x Last NRC Exam:

Question History:

                                                        -NA Memory or Fundamental Question Cognitive Level:            Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 1,4 55.43 (b) 6 Comments:

Exam Outline Cross

Reference:

Level Tier # 2 Group # 1 KIA # 005000A101 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: - Heatup/cooldown rates Importance 3.5 Question # 30 Which ONE of the following conditions would result in the core becoming uncovered earliest if a total loss of RHR occurred 120 hours after shutdown? (Assume NO operator action taken)

                           -------------------~

C&:at mid-loop with all S/G primary manways removed. Cold leg nozzle dams are installed and there have been no vent paths established. B. RCS at mid-loop with all S/G primary manways removed. No nozzle dams are installed. dams are installed and there have been no vent paths established. D. CS at mid-loop with all S/G primary manways removed. Hot leg nozzle dams are installed and the pressurizer manway has been removed. Answer: C Explanation/Justification: RCS at mid-loop with all S/G primary manways removed and hot leg nozzle dams installed provides the least water inventory. Without vent paths established, a bubble will form in the reactor vessel head causing the core to become uncovered. C is the correct answer A. Incorrect B. Incorrect C. Correct D. Incorrect Technical

References:

System Description 1.0

Proposed References to be provided: Learning Objective: 12LP-ILO-RCS001 - 13 Question Source: Bank# ---- IPEC Bank Note changes or Modified Bank # _ _ _ _ attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 5 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 005000A404 Ability to manually operate and/or monitor in the control room Controls and indications for closed cooling water pumps Importance 3.1 Question # 31 Which ONE of the following describes how the Recirculation Pump motors are cooled during an SI with blackout? A. The Recirculation Pumps are designed to run without CCW cooling for 24 hours B. CCW is re-established prior to starting the Recirculation Pumps and the Aux CCW Pumps provide additional cooling. C. The Recirculation Pumps have a backup city water cooling system that will provide city water to the Recirculation Pump coolers on low pressure D. The Recirculation Pumps have a shaft driven cooling pump that circulate CCW through the Recirculation Pump bearing cooler Answer: B Explanation/Justification: A. incorrect but plausible. Some ECCS pumps are designed to run without CCW cooling for similar periods of time. B. correct C. incorrect but plausible. RHR, HHSI and Charging Pumps have backup cooling provided from city water. D. incorrect but plausible. HHSI pumps have this feature. Technical

References:

System Description 4.1 Proposed References to be provided: None

Learning Objective: 12LP-ILO-CCW001 14 Question Source: Bank# Modified Bank Note changes or

                         #             ---       attach parent New             x Question History:              Last NRC Exam:    NA Memory or Fundamental Question Cognitive Level:      Knowledge:                          X Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 _ _ _(>-bL-)4_ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 006000A302 Ability to monitor operation of the Emergency Core Cooling System including: Pumps Importance 4.1 Question # 32 The purpose of the SI Pump Suction Low Pressure alarm is to alert the operator of loss of net positive suction head to the ... A. RHR and HHSI pumps during injection phase of SI. B. Recirc and RHR pumps during recirculation. C. HHSI pumps during low head to high head recirculation. D. HHSI pumps during injection phase of SI and recirculation. Answer: C Explanation/Justification: This alarm is activated by a switch on the supervisory panel or when Recirc Switch 6 is taken to ON. The alarm is active when RHR or Recirc pumps are supplying flow to the suction of the SI Pumps (low head to high head recirculation or Hot Leg recirculation phase) A. A. is incorrect but plausible. Candidate must recall which pumps and under which conditions the alarm is active B. is incorrect but plausible. Candidate must recall which pumps and under which condition the alarm is active C. Correct D. incorrect but plausible Candidate must recall which pumps and under which condition the alarm is active Technical

References:

ARP SBF-1 Proposed References to be provided: None

Learning Objective: Question Source: Bank# - - - - IPEC Bank Note changes or Modified Bank # __ X__ attach parent (1929) New Question History: Last NRC Exam: rJD Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 1 Group # 2 KIA # 006 Emergency Core Cooling Knowledge of the operational implications of the following 006 K5.04 concepts as they apply to the Importance 2.9 Question # 33 Which ONE of the following is the reason procedure FR-P.1, "Response to Imminent Pressurized Thermal Shock", contains less restrictive SI termination criteria than other procedures? A. To conserve water in the RWST. B. SI flow may have contributed to the RCS cooldown. C. RCS heat removal is via the steam generators and SI flow is NOT required. D. The other SI termination criteria will have already been met when FR-P.1 is entered. Answer: B Explanation/Justification: A. A. is incorrect and plausible. Any SI termination will conserve RSWT water, but that is not the reason in this case. B. is correct. FR-P1 background recovery/restoration technique section describes this as reason for early termination. C. is incorrect and plausible. These conditions are often correct for entries into FR-P.1, but not always. D. is incorrect and plausible. This statement is often true when FR-P.1 is entered but not always. Technical

References:

2-FR-P.1 Background (pg 11) Proposed References to be provided:

Learning Objective: Question Source: Bank#

                                           - -X- - IPEC    Bank 19174 Note changes or Modified Bank #             attach parent New Question History:                 Last NRC Exam:

Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b)(7) 55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 007000K101 Knowledge of the physical confleg,tiQQs and/or cause eff~ct relationships between the PRT~nd the following systems: - C~ent system . Importance 2.9 Question # 34 The Pressurizer Relief Tank (PRT) can be drained to: A. Containment Sump and Reactor Cavity Sump B. Containment Sump and Reactor Coolant Drain Tank (RCDT) C. Waste Holdup Tank and Containment Sump D. Waste Holdup Tank and Reactor Coolant Drain Tank (RCDT) Answer: C Explanation/Justification: A. Incorrect. Plausible because PRT can be drained to Containment Sump. It cannot be drained to Rx Cavity Sump B. Incorrect. Plausible because PRT can be drained to Containment Sump and to suction of RCDT pumps. A check valve prevents draining to RCDT. C. Correct D. Incorrect. Plausible because PRT can be drained to CVCS HUT (via RCDT pumps) and to suction of RCDT pumps. A check valve prevents draining to RCDT. Technical

References:

Proposed References to be provided: None Learning Objective: 12LP-ILO-RCSPZR - 6

Question Source: Bank#

                                           - -x- -   IPEC Bank 1694 Note changes or Modified Bank #             attach parent New Question History:                 Last NRC Exam:    NA Memory or Fundamental Question Cognitive Level:         Knowledge:                          X Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 _ _---->.(b-L)_3_ _ _ __ 55.43 (b) 7 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 0100002123 Conduct of Operations - Ability to perform specific system and integrated plant procedures during all modes of plant operation. Importance 4.3 Question # 35 Unit 2 is cooling down and depressurizing the RCS at the start of a refueling outage. During the process of lowering pressure from NOP to 900 psig, the following actions are performed. Select the answer that puts these actions in the proper order as pressure is lowered per 2-POP-3.3 Plant Cooldown - Hot to Cold Shutdown.

1. Monitor pressure using OPS pressure indicators PI-413K, PI-433K, or PI 443K, , 'f) ..,.., i /J \

~7J~ k s. q.,.. 175:l> PS(§ ~ f'-2~ ~ ~ ~ 4v'70 \J' Transfer PZR pressure control to manual , . "

3. Block low pressurizer pressure safety injection ~J

4. Monitor pressure using RCS hot leg pressure recorders PT-402 or PT-403 A. 2,3,4,1 B. 3,2,4,1 C. 2,3,1,4 D. 3,2,1,4 Answer: B Explanation!Justification:

A. Incorrect but plausible. It is plausible (and would work) that manual control would be used to lower pressure, but that is not what procedure specifies.

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 008000A204 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: - PRMS alarm Importance 3.3 Question # 36 The plant is operating at 100% Power. ALL of the following have occurred: Radiation Monitor R-47 is in alarm RCV-017, CCW Surge Tank Vent Valve has automatically closed. CCW Surge Tank High Level Alarm has annunciated. Which of the following events could have caused these conditions to occur and what is the appropriate procedure to address the condition? Assume all equipment intended to mitigate the event functioned as designed. A. Large tube leak in RCP Thermal Barrier HXs Go to 2 AOP-RCP-1, Reactor Coolant Pump Malfunctions B. Large tube leak in RCP Thermal Barrier HX Go to 2-AOP-LlCCW-1 Leakage into Component Cooling System C. Large tube leak in Non-regenerative HX Go to 2-AOP-CVCS-1, CVCS Malfunctions D. Large tube leak in Non-regenerative HX Go to 2-AOP-LlCCW-1 Leakage into Component Cooling System

Answer: D ExplanationlJustification: The only auto closure signal for RCV-017 is high activity on Radiation Monitor R 47. A. Incorrect. Plausible because a large leak in the thermal barrier heat exchanger will cause leakage into CCW but will cause FCV-625 to auto close. Closing 625 will stop the leakage and RCV-017 will not auto close. In addition AOP-RCP-1 is not the correct procedure to address this condition. B. Incorrect. Plausible because a large leak in the thermal barrier heat exchanger will cause leakage into CCW but will cause FCV-625 to auto close. Closing 625 will stop the leakage and RCV-017 will not auto close. AOP-LlCCW-1 is the correct procedure to address this condition. C. Incorrect. Plausible because a large leak in the Non-regenerative heat exchanger will cause leakage into CCW and may cause RCV-017 to auto close. In addition AOP-CVCS-1 is not the correct procedure to address this condition. D. Correct. A large leak in the Non-regenerative heat exchanger will cause leakage into CCW and may cause RCV-017 to auto close. AOP-LlCCW 1 is the correct procedure to address this condition. Technical

References:

System Description 4.1 Proposed References to be provided: None Learning Objective: 12LP-ILO-CCW001 - 9 Question Source: Bank# IPEC Bank

                                             -----          Note changes or Modified Bank #
                                             - -x- -        attach parent 18782 New Question History:                   Last NRC Exam:         NA Memory or Fundamental Question Cognitive Level:           Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _--->('-'b)<-4_ _ _ __ 55.43 (b) 7

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 010000K603 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PC5: - PZR sprays and heaters Importance 3.2 Question # 37 Offsite power is lost without 51 actuation.

The control room operators verify a reactor trip and a turbine trip.
They determine that the EOG's have energized the 480 V AC busses.
All appropriate loads have sequenced on.
While ensuring that the RC5 stabilizes at no-load conditions, an operator observes that PZR pressure is 2100 psig and slowly lowering.
PZR PORV'S and spray valves are closed.
PZR level has risen from 19% to 25%.

What corrective action, if any, should be taken? A. No operator action is necessary. B. Manually actuate 51. C. Maximize charging flow. O. Reset the PZR Backup heaters. Answer: 0 Explanation!Justification: A. Incorrect but plausible because an operator may not know when pressure will start to recover and think heaters are available B. Incorrect but plausible because an operator may interpret data as indicating 51 needed due to pressure C. Incorrect but plausible since an operator may assume that charging is used to raise pressure

D. Correct. Loss of off-site power blocks auto closure of backup heater breakers. This must be reset to allow breakers to close. Technical

References:

2-E-O Reactor Trip or Safety Injection Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPEOO - 4 Question Source: Bank# x

                                             --------     IPEC Bank 3683 Note changes or Modified Bank #     -----        attach parent New Question History:                  Last NRC Exam:       -NA Memory or Fundamental Question Cognitive Level:          Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: Comments:

Exam Outline Cross

Reference:

Level RO SRO Tier # 2 Group # 1 KIA # 012000K502 31~ Knowledge of the operational

                         ~t>                        implications of the following
                   ~~

concepts as they apply to the RPS: - Power density

Importance 3.1 Questio # 38 The pia t is at 80% power during a power ascension. Assumi g that RCS and flux distribution parameters remain on program!target, as power is raise to 100%, how will the over-temperature (OT) and over-power (OP) differential temperature (DT) Reactor Protection setpoints change? OTAT setpoint OPAT setpoint A. increase stay the same B. decrease decrease C. decrease stay the same D. stay the same increase Answer: C Explanation!Justification: A. Incorrect. Plausible because the OPOT setpoint does remain the same, and the applicant may confuse the fact that the OTOT setpoint decreases with the term 'increase' meaning the actual value is closer to setpoint. B. Incorrect. The OTOT setpoint does decrease. The OPOT setpoint never increases from its nominal value. It will, however decrease if T -avg deviates above its nominal 100% power program value. Since T-avg at 80% power is less than 100% power, the OPOT setpoint will be at its nominal full power value and thus, will not change from 80 to 100% power assuming T-avg stays on program.

C. Correct. Since T -avg at 80% power is less than 100% power, the OPOT setpoint will be at its nominal full power value and thus, will not change from 80 to 100% power assuming T-avg stays on program. The OTOT setpoint, on the other hand can increase or decrease from its nominal value. Since program T-avg will increase approximately 5 more degrees, the trip setpoint will become more limiting, decreasing to its nominal full power value. O. Incorrect. OTOT decreases as referenced above in C. The OPOT setpoint never increases from its nominal value. It will, however decrease if T-avg deviates above its nominal 100% power program value Technical

References:

Technical Specifications 3.3.1 Proposed References to be provided: None Learning Objective: 12LP-ILO-ICRXP - 3 Question Source: Bank# Modified Bank # X IPEC Bank Note changes or _____ attach parent New Wolf Creek Question History: Last NRC Exam: 2009 Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: x 10 CFR Part 55 Content: 55.41 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 't------ r022000K404

                                                                      ---=
                                                                         ---"-<::::::~:~-"~-,,

KIA #

                                                   , Knowledge of CCS design " \

feature(s) and/or interlock(s) ) which provide for the following: /

                                                     ~ Cooling of control rod drive//
                                                                                       ./

motors Importance 2.8 Question # 39 Unit 2 operators are performing 2-ES-O.2, Natural Circulation Cooldown.

2-ES-O.2 has been entered because offsite power had been lost.
The EDGs started and energized the AC emergency buses.
The CRDM cooling fans cannot be manually loaded onto the AC emergency buses.
Condensate Storage Tank water inventory is adequate for the cooldown.

Which ONE of the following describes HOW the inoperability of the CRDM fans affects the cool down and depressurization? A. It has no effect because the amount of RCS heat removed by running the CRDM fans is insignificant compared to the heat removed by steaming the secondary plant. ~:: The RCS Cooldown rate must be reduced since the upper head can not be adequately cooled. Transition to 2-ES-O.3, Natural Circulation Cooldown with Steam Void in Vessel, will be required because cooldown and depressurization will cause formation of a void in the upper head area. D. Less subcooling should be maintained to enhance the cooldown of the upper head area, which reduces the formation of voids. Answer: B Explanation!Justification:

A. Incorrect - Does NOT address the issue of a reduced upper head cooldown rate. B. Correct - ES-0.2 will have restriction without cooling fans. but transition to ES-0.3 is not required. C. Incorrect - Transition to 2-0HP-4023-ES-0-3, Natural Circulation Cooldown with Steam Void in Vessel, is NOT required given conditions which do NOT warrant an increased cooldown rate on natural circulation (Le., CST inventory adequate for cooldown). D. Incorrect - The absence of the CRDM fans requires a greater RCS subcooling Technical

References:

2-ES-0.2 Natural Circ Cooldown Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPS02 - 5 Question Source: Bank#

                                              - - -x- - - IPEC Bank Note changes or Modified Bank #     - - - - - attach parent New DC Cook Question History:                   Last NRC Exam:         2002 Memory or Fundamental Question Cognitive Level:           Knowledge:                                  x Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 (b) 5 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 013000K601 Knowledge of the effect of a loss or malfunction of the following will have on the ESFAS: - Sensors and detectors Importance 2.7 Question # 40 Given the following:

The plant is at 100% power.
All control systems are in their normal alignments.
Pressurizer Pressure Transmitter PT-455 has failed LOW.
All actions have been taken to remove the transmitter from service in accordance with 2-AOP-INST-1, Instrument or Controller Failures.

Which ONE of the following describes the logic required from the remaining operable pressurizer pressure channels to initiate (1) a Low Pressurizer Pressure Reactor Trip, and (2) a Low Pressurizer Pressure Safety Injection actuation? A. (1) 1 out of 2 (2) 1 out of 3 B. (1) 1 out of 3 (2) 1 out of 2 C. (1) 1 out of 2 (2) 1 out of 2 D. (1) 1 out of 3 (2) 1 out of 3 Answer: B ExplanationlJustification: Pressurizer Low Pressure Reactor Trip is a 2 out of 4 logic. Pressurizer Low Pressure SI is a 2 out of 3 logic A. Incorrect. Opposite of actual

8. Correct C. Incorrect. Reactor Trip receives inputs from 4 channels D. Incorrect. Safety Injection receives input from 3 channels Technical

References:

                     -'=~9l~Qla.9.J251 ~~~ _______._._.~

Proposed References to be provided: None

                                          -~----
                                          --.---.,-.~-.----

Learning Objective: 12LP-ILO-ESS001 - 4

                                          -"--~------~-----------~----~--.--.---

12LP-ILO-ICRXP - 3 Question Source: Bank# IPEC Bank Note changes or Modified Bank # - .. -.- .....* -...... -~ attach parent New x Question History: Last 2 NRC Exams at IPEC: NA Memory or Fundamental Question Cognitive Level: Knowledge: x Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 5 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier# 2 Group # 1 KIA # 022000K301 Knowledge of the effect that a loss or malfunction of the CC5 will have on the following: Containment equipment subject to damage by high or low temperature, humidity, and pressure Importance 2.9 Question # 41 Given:

A small break LOCA occurred approximately 45 minutes ago.
Containment pressure peaked at 8 psig.
Containment pressure subsequently decreases to 2.5 psig.
Containment radiation peaked at 15 Rlhr on R-25 and 26
The crew is performing an 51 flow reduction using E8-1.2, Post-LOCA Cooldown and Depressurization.
One 51 pump has been secured using Adverse Containment Values.
A second pump cannot be secured at this time.
5ubcooling remains constant.

Which of the following is true regarding securing subsequent 81 Pumps? A. The pump cannot be secured using normal values; once flow reduction is started using adverse containment values,-&t;lbsequent flow reduction actiens continue using adverse containment values. B. The pump cannot be secured using normal values; since pump was evaluated using adverse containment values; it must be secured using adverse containment values. ~Bseguellt actions cOiltillue using normal I

      ~                                                                           t C. The Pump can be secured using normal containment values, when adverse containment pressure and radiation values return to normal, l'lQC.tRal eOI Itainment values are used.

D. The pump can be secured using normal containment values, when* containment pressure decreases to acceptable ranger+termal eOfltainment

      \1"'1**t::>.,...,
      ~u.;;>~u.         ..~*-

Answer: 0 Explanation!Justification: Two conditions result in adverse containment conditions, VC pressure and radiation. If pressure decreases to less than the adverse value, the crew should use normal values. If radiation levels decrease to normal values, the crew should continue using adverse containment values. A. Incorrect. Plausible because as discussed above, one conditions result in returning to normal values (pressure) and one condition (radiation) does not. It is reasonable that once the flow reduction using adverse values is initiated, it should continue using adverse values. B. Incorrect. Plausible because as discussed above, one conditions result in returning to normal values (pressure) and one condition (radiation) does not. It is reasonable that once the pump itself was evaluated using adverse values is initiated; it should continue using adverse values. C. Incorrect. If containment radiation levels return to normal range, adverse values are used until engineering evaluation allows use of normal values .. D. Correct Technical

References:

OAP-12 Proposed References to be provided: NA Learning Objective: 12LP-ILO-EOPROU - 21 Question Source: Bank# IPEC Bank Note changes or Modified Bank # - - - - attach pa rent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier# 2 Group # 1 KIA # 026000K201 Knowledge of bus power supplies to the following: Containment cooling fans Importance 3.0 Question # 42 The following conditions exist:

Reactor Trip and Safety Injection have actuated.
Main Steam Line Isolation and Containment Isolation Phase B have actuated.
480 volt vital bus 5A is deenergized due to a fault.

Which of the following describes the equipment available to reduce containment pressure? A. 22 CS pump, 23, 24, 25 Containment Fan Cooler Units B. 21 CS pump, 21, 24, 25 Containment Fan Cooler Units C. 21 CS pump, 22, 24, 25 Containment Fan Cooler Units D. 22 CS pumps, 21, 24, 25 Containment Fan Cooler Units Answer: A Explanation/Justification: IPEC has 3 Safeguards buses (5A, 2A-3A, and 6A) Any 2 safeguards buses satisfy minimum safeguard power requirements. Equipment is distributed among the safeguards buses. Candidates must know the what equipment is power from what safeguards bus. Bus 5A is the any 480 V bus with 2 FCUs (Buses 2A and 3A each haveJ FCU) A. Correct B. Incorrect. Plausible because the 3 FCUs are correct. The power supply to 21 CS pump is 5A. C. Incorrect. 21 CS Pump and 22 FCU are power from 5A. D. Incorrect. 22 CS Pump and 22 FCU are power from 5A Technical

References:

2-AOP480V-1 Att 5

Proposed References to be provided: Learning Objective: 12LP-ILO-ESS001 - 3 Question Source: Bank#

                                            - -X - -   IPEC Bank 17513 Note changes or Modified Bank #               attach parent New Question History:                 Last NRC Exam:      NA Memory or Fundamental Question Cognitive Level:         Knowledge:                            x Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 (b) 7 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 039000K405 Knowledge of MRSS design feature(s) and/or interlock(s) which provide for the following:

                                                                    - Automatic isolation of steam line Importance                           3.7 Question #       43 A fault occurred on 23 SG inside containment, and a high steam flow was sensed on 23 Main Steam Line. 23 SG pressure is 500 psig and decreasing. Which ONE of the following is correct?

A. All Main Steam Isolation Valves will close immediately. B. Only the Main Steam Isolation Valve for the 23 Steam Generator will close. C. All Main Steam Isolation Valves will close when 23 Steam Generator Pressure is 155 psid below the pressure in the other Steam Generators. D. None of the Main Steam Isolation Valves will be immediately affected. Answer: D Explanation/Ju stification: A. High flow conditions are required in two of the four steam generators concurrent with low Tave or low Pressure B. Individual MSIVs close when the control switch is placed in TRIP or air is lost to that MSIV for an extended period of time C. Steam Line differential pressure causes a safety injection signal not a MSIV isolation signal D. Correct answer. Drawing 241685 Technical

References:

                      ~D_r_a_w_in.......g,-2_4_1_6_8_5_ _ _ _ _ _ __

Proposed References to be provided: -=N-=o;;...:.n.;...:.e'--_ _ _ _ _ _ _ _ _ __ Learning Objective: 12LP-ILO-ESS001 - 5

Question Source: Bank#

                                           - -x- -   IPEC Bank 5149 Note changes or Modified Bank #             attach parent New Question History:                 Last NRC Exam:    NA Memory or Fundamental Question Cognitive Level:         Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _---->(--I.b)'--7_ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 059000K302 Knowledge of the effect that a loss or malfunction of the MFW System will have on the following: - AFW System Importance 3.6 Question # 44 The plant was operating at 30% power during a power ascension with 21 and'22 MBFP operating in AUTO when 21 MBFP trips on low auto stop oil pressure. When can 21 and 23 AFW pumps be secured and not auto-restart with the switches in auto?

~~

A. After 21 MBFP Trip Switch is placed in trip.

/'

B. Any time adequate MFW flow exists. C. Any time SG level is greater than 9% in all SGs. D. When 21 MBFP auto stop oil pressure returns to > 25 psig. Answer: A Explanation/Justification: A. Correct (2-AOP-FW-1 Step 102) B. Incorrect. Plausible because adequate feed flow would maintain SG level above the auto start setpoint C. Incorrect. Plausible because 9% is the auto start setpoint; however, SG level will not decrease to 9% before the AFW pumps auto start on MBFP trip. D. Incorrect. Plausible because 28 psig is the auto stop oil setpoint that will cause the MBFP to trip, so the pump trip cannot be reset at 25 psig. Technical

References:

2-AOP-FW-1

                                          ~------------------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-MFW001 - 9

Question Source: Bank #

                                           - - - - IPEC Bank Note changes or Modified Bank #   --~-

attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _---->.(b.c...L.)_7_ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 061000A101 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the AFW System controls including: - S/G level Importance 3.9 Question # 45 Unit 2 was operating at 100% power when 22 RCP tripped due to a fault. All equipment operated as designed. Which of the following describes how 21 and 22 SG levels will respond to this event and why? A. 22 SG level will increase at a faster rate than the 21 SG. AFW flow is greater to 22 SG because it is at a lower pressure

8. 22 SG level will increase at a slower rate than the 21 SG. AFW flow is lower to 22 SG because it is at higher pressure C. 22 SG level will increase at a slower rate than the 21 SG because it is steaming at a higher rate.

D. 22 SG level will increase at a faster rate than the 21 SG because it is steaming at a lower rate. Answer: D ExplanationlJustification: A. Incorrect: Plausible because 22 SG level will increase at a faster rate, but AFW flow is automatically controlled at approximately 200 gpm to each SG

8. Incorrect: Plausible because candidate may confuse SG pressure response on trip of an RCP.

C. Incorrect: Plausible because 22 SG is one of the supplies to the TDAFW pump. Candidates may believe that this will cause the level increase to be slower. D. Correct: With non-return check valves in the steam lines, 22 SG will steam at a much lower rate even though it is supplying 22 AFW pump turbine. Technical

References:

Proposed References to be provided: None Learning Objective: 12LP-ILO-RCSRCP - 15 Question Source: Bank # - - - - - IPEC Bank Note changes or Modified Bank # - - - - attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 14 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 062000A101 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the A.C. Distribution System controls including: Significance of D/G load limits Importance 3.4 Question # 46 While transferring the plant to Cold Leg Recirculation, it is noted that #21 EDG loading is currently 1650 KW. The next step to be performed in ES-1.3 is to place Safety Injection Recirculation Switch 4 to "ON", which will start 21 ' Recirculation Pump t£99 t(W1. What is the major consideration for continuing with this step of the procedure? A. Load must be removed from the bus prior to starting the pump to prevent exceeding the maximum short time (2 hr) load rating of the EDG. B. Load must be removed from the bus prior to starting the pump to prevent exceeding the continuous EDG load rating. C. Starting this pump is allowed, but EDG load will be limited to the short time (2 hr) load limit after this pump is placed in service. D. 22 Recirculation Pump is started, per procedure, to prevent an overload condition on the EDG. Answer: C Explanation/Justification:

This question requires the candidate to have the following knowledge:

1. what the continuous and 2 hour limits on EDGs are

2. Have a rough idea of KW of recirc pumps. Note that the KW values picked would be valid for any pump that would be run in ES-1.3. All safeguards motors are >100 KW and < 450 KW. Recirc pumps are 299 KW.

3. Procedure actions of ES-1.3 (we do not formally address EDG load).

Additionally, the candidate has to piece together the information to come to conclusion that starting this pump will put us over the normal limit, but within the short term limit and that this is OK. A. Incorrect but plausible. It is possible that a candidate may believe that the short term limit would be exceeded. B. Incorrect but plausible. The continuous load rating will be exceeded, but we do not have to shed load. C. Correct. Per 2-sop-27.3.1.1 caution at step 4.2.10, this load is allowed for 2 hours in a 24 hour period. D. Incorrect but plausible. It is reasonable for a candidate to assume that 22 pump would be used, but the procedure does not have steps to check for load prior to stating 21 pump with recirc switch 4. Technical

References:

2-S0P-27.3.1.1 Proposed References to be provided:

                                             -.:N_._o.::...:n..:...;e:..-_ _ _ _ _ _ _ _ _ ___

Learning Objective: 12LP-ILO-EDSEDG - 8 Question Source: Bank#

                                                            ---           x -- IPEC Bank 3035 Note changes or Modified Bank #                _____ attach parent New Question History:                       Last NRC Exam:                         NA Memory or Fundamental Question Cognitive Level:               Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _---->(---<b)'---7_ _ _ __ 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 062000A201 Ability to (a) predict the impacts of the following malfunctions or operations on the A.C. Distribution System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Types of loads that, if de energized, would degrade or hinder plant operation Importance 3.4 Question # 47 The plant is at power with the following conditions present:

100% power
All control systems are in automatic
Tave - 565 degrees F
RCS press - 2235 psig
No equipment out of service
Pressure channel 3 is in control
Pressure channel 2 is the alarm channel Which one of the following correctly describes one effect of losing 22 instrument bus?

A. PORV 455C is prevented from Automatically opening B. PORV 456 is prevented from automatically opening C. PORV 456 Block Valve (536) will not automatically open. D. BOTH PORVs are prevented from automatically opening. Answer: B

Explanation/Justification: A. Incorrect. Plausible because the candidate must recall that PT-456 (Channel 2) is powered from 22 IB and that it cannot be in control. Thus it cannot affect PORV 455C. B. Correct. Candidate must recall that PT-456 (Channel 2) is powered from 22 IB and loss of PT-456 will prevent PORV 456 from opening. C. Incorrect. Plausible. The candidate must recall which pressure channels provide opening signals to the PORV Block Valves. (PT-457 opens Block Valve 536 and PT-474 opens Block Valve 535). D. Incorrect. Plausible. The candidate must recall which pressure channels provide opening signals to the PORV Block Valves. (PT-457 opens Block Valve 536 and PT-474 opens Block Valve 535). Technical

References:

System Description 1.0 Proposed References to be provided: None Learning Objective: 12LP-ILO*RCSPZR - 9 Question Source: Bank# IPEC Bank

                                             -------      Note changes or Modified Bank #                 attach parent New                     x Question History:                   Last NRC Exam:
                                                      -NA- - -

Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 7

                                                    -----~~----------

55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 063000A403 A ility to manually operate d/or monitor in the control room: - Battery discharge rate

                                            ~f Importance         3.0 Question #      48 A 345KV fault leads to     trip of Unit 2 from 100% power.

Due to the electrical tf. nsient, 21 Battery Charger trips and cannot be restarted. Based on these con itions and equipment design criteria, what is the expected plant response? A. 21 Battery i designed to ensure voltage will remain above a predeter ned acceptable value for 2 hours. After voltage drops below this level, 80V switchgear and safeguards relays powered from 21 DC~ will switch 0 an alternate source. B. 21 Batte is designed to ensure voltage will remain above a predeter ined acceptable value for 2 hours. After voltage drops below this level, 480V switchgear powered from 21 DC will switch to an alternate source. C. 21 Battery designed to ensure voltage will remain above a predeter ined acceptable value for 4 hours. After voltage drops below this level, 80V switchgear and safeguards relays powered from 21 DC will switch to an alternate source. D. 21 Battery s designed to ensure voltage will remain above a predeter ned acceptable value for 4 hours. After voltage drops below this level 80V switchgear powered from 21 DC will switch to an alternate source. Answer: B ExplanationlJustification:

This question tests two knowledge areas. One being how long batteries are rated for (2 hours per T.S. Basis). The other being what automatically swaps to alternate power on a loss of DC, which is only switchgear. A. Incorrect but plausible. The safeguards relays do not switch power source, but it a plausible answer because switchgear does. B. Correct C. Incorrect but plausible. There are many 4 hour ratings. It is plausible that a candidate could think 4 hours is correct. Also see above for including safeguards relays. D. Incorrect but plausible. See above for 4 hours vs. 2 hours. Technical

References:

Tech Spec 3.8.4 Basis Proposed References to be provided: _N_o-'-n_e-'--_ _ _ _ _ _ _ _ _ _ __ Learning Objective: 12LP-ILO-EDS03 - 12 Question Source: Bank # - - - - IPEC Bank Note changes or Modified Bank # attach parent New x Question History: Last NRC Exam: NA

                                                            ~c.,.;- __

Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 _ _---'(--'b)'--7_ _ _ _~ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier# 2 Group # 1 KIA # 064000K102 Knowledge of the physical connections and/or cause effect relationships between the ED/G System and the following systems: - ED/G cooling water system Importance 3.1 Question # 49 What would happen if the Jacket Water Pump on a Emergency Diesel Generator had a broken shaft and the Emergency Diesel Generator received an AUTO start signal, with no operator action. A. The Emergency Diesel Generator would run until it overheated, then low oil pressure would trip the 86 device B. The Emergency Diesel Generator would start and continue to run, but the field would not 'flash' so there would be no generator output C. The Emergency Diesel Generator would start but only run for about 13 seconds, then the 86 would trip D. Without jacket water pressure the Emergency Diesel Generator would start, run for 2 minutes and shut down normally Answer: C Explanation/Justification: A. Incorrect. Plausible because the EDG would overheat without cooling. The engine start failure would trip the diesel before this occurrs. B. Incorrect. Plausible because the EDG would overheat without cooling. The engine start failure would trip the diesel before this occurs. C. Correct D. Incorrect. Plausible because the engine would start; howver it would not run for 2 minutes nor would it shutdown normally. Technical

References:

System Description 27.3 Proposed References to be provided: None

Learning Objective: 12LP-ILO-EDSEDG - 10 Question Source: Bank#

                                           - -x- -   IPEC Bank 6756 Note changes or Modified Bank #             attach parent New Question History:                 Last NRC Exam:     NA Memory or Fundamental Question Cognitive Level:         Knowledge:                          X Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 _ _--->.{b-L}_7_ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier# 2 Group # 1 KIA # 073000K301 Knowledge of the effect that a loss or malfunction of the PRM System will have on the following: - Radioactive effluent releases Importance 3.6 Question # 50 A liquid release is in progress. Power is lost to R-54, Liquid Radiation Monitor. Assuming all components functioned as designed, what is the status of the Waste Distillate System? I Waste Dist WDTP Disch WDTP Recirc WDTP Common I Trans Pump Valve Valve Disch Valve I A. Tripped Open Closed Closed lB. Running Closed Open Open I

C. Tripped Closed Open Closed

! D. Running Open Closed Open' Answer: C Explanation/Justification: A. Incorrect. Plausible because the common discharge valve (LWR-701) will close and the pump trip which would stop the leak; however, the pump discharge valve will close and the recirc valve will open. B. Incorrect. Plausible because closing the pump discharge valve and opening the pump recirculation valve will stop the release and continue mixing of the Distillate Storage Tank contents.

C. Correct D. Incorrect. Candidate may not recognize that a loss of power to the radiation monitor results in the same automatic actions as a high radiation level alarm. Technical

References:

2-S0P-12.3.3

                                           ~~~~~~-----------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-RMS001 - 2 Question Source: Bank# IPEC Bank Note changes or Modified Bank # _____ attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 13 55.43 (b) 4 Comments:

Exam Outline Cross

Reference:

Level RO Tier# 2 Group # 1 KIA # 076000A201 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Loss of SWS Importance 3.5 Question # 51 Unit 2 is in Mode 4 and has experienced an unexpected rise in containment sump level. Per 2-FR-Z.2, Response to Containment Flooding, which of the following systems could cause containment water level to exceed 49.5 feet? A. Reactor Coolant System B. Chemical and Volume Control System C. Residual Heat Removal System D. Service Water Answer~bh ~ VIL R1J .u,.,)&JJ~~ /J'It-;:U.,., Explanafr nlJustification ~ . - - fI -71' L.f This question is challenging because the plant is in Mode 4. A. Incorrect based on background document for FR-Z2, which states that the level for determining flooded VC is based on all of RCS, RWST, and CST has entered VC .. Plausible since RCS inventory is potentially much greater than in Mode 1. B. Incorrect per FR-Z3 background. Plausible since Mode 4 CVCS lineups could put more water in VC than Mode 1. C. Incorrect per FR-Z3 background document. Plausible since RHR being in service could put all of RCS plus RWST in VC.

D. Correct Technical

References:

2-FR-Z.2 Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPFZ2 - 1 Bank# Question Source:

                                           - -X- -   IPEC Bank Note changes or Modified Bank #   ----      attach  parent New McGuire Question History:                 Last NRC Exam:    2005 Memory or Fundamental Question Cognitive Level:         Knowledge:

Comprehension or Analysis: x 10 CFR Part 55 Content: 55.41 (b) 4 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 0760002132

                                               ~ J      ----------------------

Conduct of Operations - Ability P ..v-~

                                                      . to explain and apply all system limtts and precautio:V" I.fA Question #       52    .. /
                           ~         Importance                3.8
                                                                        /
                                                                           <;~

Which ONE of the fo~wipg describes \l1e stgfling duty requirements for the Service Water Pumps~ ~ ~ ? ( A. Two consecutive starts are allowed, a third restart is allowed after pump has been idle for 30 minutes. B. After the first start, a second restart is allowed after the pump has been running for 10 minutes. C. ~!5 ~ecutive starts allowe1~a third estart is allowe?~ aft" the pump has been idle for 60 minutes. C fI:.e !14tv~ ~V~ t/.i-tLS#YJ~ 5~ After the first start, second restart aft9t:. iS1110wed 8~er the pump has been idle for : minutes. ~ ~ Explanation/Justification: ~S~i4 (Fa ~ oV()/ A. Incorrect. Plausible because third restart is allowed if pump has RUN for 30 minutes. B. Incorrect. Plausible because two consecutive restarts are allowed. It is not a REQUIREMENT to run the pump for any time for the second start. C. Correct D. Incorrect. Plausible two consecutive restarts are allowed. It is not a REQUIREMENT for the pump to be idle for any time for the second start. Technical

References:

2-S0P-24.1

                                            ~------------------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-SW001 - 6

Question Source: Bank# IPEC Bank Note changes or Modified Bank # attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 _ _---l.(;;;;.lb)'--=8_ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier# 2 Group # 1 KIA # 078000K201 Knowledge of bus power supplies to the following: Instrument air compressor Importance 2.7 Question # S3 Given: A unit trip has occurred from 2S% power A fault occurred on the Station Aux Transformer 21 Diesel Generator started but the output breaker failed to automatically close. Which of the following correctly identifies the air compressors available to supply instrument and station air? A. 11 and 12 SAC (Centac), 21 & 22 Instrument Air Compressors B. 21 & 22 Instrument Air Compressors, Station Air Compressor C. 11 and 12 SAC (Centac), 22 Instrument Air Compressor D. 22 Instrument Air Compressor, Station Air Compressor Answer: C Explanation/Justification: 11 SAC (Centac air compressor) is power Unit 1 buses 11 SA 1 12 SAC (Centac air compressor) is power Unit 1 buses 12SA2 21 Instrument Air compressor is powered from MCC29A (from Bus SA) 22 Instrument Air compressor is powered from MCC24A (from Bus 2A) Station Air Compressor is powered from bus SA 21 EDG supplies bus SA. Without bus SA, 21 Instrument Air and the Station Air Compressors are not available. A. Incorrect B. Incorrect C. Correct D. Incorrect Technical

References:

COL-29.3 and 27.1.S

Proposed References to be provided: Learning Objective: 12LP-ILO-SA01 - 6 12LP-ILO-SA01 - 5 Question Source: Bank# - - - IPEC Bank Note changes or Modified Bank # _ _ _ _ attach parent New x Question History: Last NRC Exam:

                                                   -NA- - -

Memory or Fundamental Question Cognitive Level: Knowledge: x Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 078000A301 Ability to monitor automatic operation of the lAS, including:

                                                     - Air pressure Importance          3.1 Question #       54 Given the following plant conditions:

Plant is in cold shutdown with RCS depressurized RHR cooling is in service, vessel level is at 68' Vessel head de-tensioning in progress An instrument air line ruptures in the AFB: IA header pressure is 65 psig and decreasing Crew enters 2-AOP-AIR-1, Air System Malfunctions Which ONE of the following statements is correct in regards to the status of the RCS? A. No effect(s) on the RCS given the break location; a check valve in the IA supply line will effectively isolate the AFB from the rest of the station. B. RCS level will increase without operator action due to letdown isolation and charging pump speed increase. C. RCS level will decrease because charging line isolation valves 204A & B will fail closed and HCV-133, RHR letdown isolation valve, fails open. D. RCS temperature is going to increase due to the isolation of CCW to the RHR heat exchangers. Answer: B Expl anation/Justification: A. Incorrect. Plausible because a pressure regulating valve could perform this function; howver, no such valve exists in the supply line to the AFB. B. Correct. Candidate should know the fail position of major valves/components.

C. Incorrect. Candidate should know the fail position of major valves/components. 204A & B fail open, HCV-133 fails closed. D. Incorrect. Candidate should know the fail position of major valves/components. 822 valves are MOVs. Technical

References:

2-AOP-AIR-1

                                       ~---------------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-SA01 -14 Question Source: Bank#

                                           ---    X --    IPEC Bank - 5491 Note changes or Modified Bank #                  attach parent New Question History:                 Last NRC Exam:          NA Memory or Fundamental Question Cognitive Level:         Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 4 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 1 KIA # 103000K406 Knowledge of Containment System design feature(s) and/or interlock(s) which provide for the following: Containment isolation system Importance 3.1 Question # 55 / Which ONE of the following will result in 8f1 Main Feedwater Isolation with the plant operating at 75% power? .

~ A. Pressure transmitter 419A for SG 21 has failed high and Pressure 7~        Transmitter 449A for SG 24 has failed HIGH.

B. Level transmitter for 437A for SG 23 has failed high and Level transmitter 447C for SG 24 has failed HIGH. C. Containment pressure transmitter 948A has failed HIGH and Containment pressure transmitter 949B has failed HIGH. D. Pressurizer pressure transmitter 456 (Channel 2) has failed LOW and Pressurizer pressure transmitter 474 (Channel 4) has failed LOW. Answer: D ExplanationlJustification: A. Incorrect. Plausible because condition requires evaluation to see if delta PSI signal will be generated. An SI signal will not be generated. B. Incorrect. Plausible because High SG level causes feedwater isolation however the coincidence is not correct. C. Incorrect. Plausible because Containment Pressure will cause a safety injection. The coincidence is 2 of 3 on the 948A-C instruments. The 949A-C transmitters are used for Containment Spray. D. Correct. Pressurizer Low Pressure Reactor trip with low Tavg Technical

References:

System Description 21 Proposed References to be provided: _N_o~n__e,-_ _ _ _ _ _ _ _ _ __ Learning Objective: 12LP-ILO-ESS001 - 5

Question Source: Bank # IPEC Bank Note changes or Modified Bank # _ _X_ _ attach parent 4232 New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: x Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier# 2 Group # 2 KJA# 002000A201 Ability to (a) predict the impacts of the following malfunctions or operations on the RCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Loss of coolant inventory Importance 4.3 Question # 56 Given:

A SBLOCA occurred leading to a reactor trip and safety injection.
SI termination criteria were satisfied and the crew is evaluating if letdown can be re-established when it is observed that PZR level is 16% and slowly lowering.
21 Charging Pump is running in manual at maximum speed.

Based on these conditions what is the appropriate action to take? A. Manually start SI Pumps as necessary to restore level and go to E-1, Loss of Primary or Secondary Coolant. B. Manually start SI Pumps as necessary to restore level and go to ES-1.2. Post-LOCA Cooldown and Depressurization. C. Manually actuate SI and go to E-O, Reactor Trip or Safety Injection. D. Manually actuate SI and go to E-1, Loss of Primary or Secondary Coolant. Answer: A Explanation/Justification: A. Correct per the foldout of ES-1.1

B. Incorrect but plausible. The team will be eventually going to ES-1.2 to perform an SI reduction, but E-1 is entered first. C. Incorrect but plausible. This action is the action for similar indications in other EOPs (e.g. ES-O.1) D. Incorrect but plausible. This action is very similar to what is specified in the procedure and in fact would put the team in the identical conditions that would have existed if E-1 was initially entered. Technical

References:

2-ES-1.1 Foldout

                                          ~-----------~~--~--------

Proposed References to be provided: None

                                          -----------~------~=-~----

Learning Objective: 12LP-ILO-EOPS11 - 2 Question Source: Bank # IPEC Bank Modified Bank # X New McGuire Question History: Last NRC Exam: 2003 Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 2 KIA # 015000K201 Knowledge of bus power supplies to the following: - NIS channels, components, and interconnections Importance 3.3 Question # 57 The following conditions exist:

A plant startup is in progress.
Reactor power is currently 7%.
A loss of Instrument Bus 22 occurs.

Which ONE of the following describes the effect on the plant? A. Source Range instruments energize prematurely. B. Reactor trips due to loss of one Source Range instrument. C. Reactor trips due to loss of one Intermediate Range instrument. D. Intermediate Range high flux reactor trip will NOT actuate if required. Answer: C ExplanationlJustification: A. Incorrect. The logic to re-energize the SR Nls is 2 of 2 IR < P-6 setpoint. Plausible the candidate must remember that both IR Nls must be < P-6. B. Incorrect. With the P-6 block still in tact, the SR tips are bypassed. Plausible because N31 is powered from IB 21 and if operating in the SR, loss of power to N-31 would cause a Rx Trip. C. Correct. IB 21 supplies one channel of IR NIS (N-35). Loss of power to the channel will result in loss of power to protection bistables. The IR trip is a 1 of 2 coincidence; thus causing the trip. D. Incorrect. Plausible because the candidate must understand that both control and instrument power are lost to the IR channel when IB-21 is de energized. Furthermore the candidate must understand what effect de energizing bistable relays has on Reactor Protection.

Technical

References:

2-S0P-13.1

                                       ~~--~~------------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-ICESC - 7 12LP-ILO-ICESC - 9 Question Source: Bank# IPEC Bank

                                           -----       Note changes or Modified Bank #   _____ attach parent New                    x Question History:                 Last NRC Exam:
                                                    -NA- - -

Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 _ _----'-(b:.;....L)__7_~~_ _ 55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 2 KIA # 011000A102 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR LCS controls including: - Charging and letdown flows Importance 3.3 Question # 58 The following conditions exist:

The RCS is being taken solid during a cooldown.
Cooldown rate is approximately 50°F/hr
24 RCP is in operation
Actual Pressurizer Level is 90% and slowly rising
Pressurizer Pressure is 350 psig and stable At 95% Pressurizer level the cooldown rate is reduced to 30°F/hr. How will this effect Pressurizer fill rate and what actions must be taken?

A. Pressurizer fill rate will increase. Decrease charging pump speed B. Pressurizer fill rate will increase Reduce PCV-135 auto setpoint C. Pressurizer fill rate will decrease Increase charging pump speed D. Pressurizer fill rate will decrease Increase PCV-135 auto setpoint Answer: A Expla nationlJustification:

A. Correct. Reducing cooldown rate will reduce rate of contraction of coolant and increase the fill rate. Charging flow must be reduced to remain within procedure guidelines. B. Incorrect. Plausible because reducing cooldown rate will reduce rate of contraction of coolant and increase the fill rate. Reducing PCV-135 setpoint will increase letdown flow; however it will also reduce RCS pressure needed to maintain RCS in operation. The procedure directs . only adjustment in charging flow. C. Incorrect. Plausible because the candidate must understand the effects of changing cool down rates on the fill rate in the pressurizer. In addition. increasing charging flow would be appropriate if the fill rate was reduced;* ~ however it is not correct since fill rate will actually be increased. D. Incorrect. Plausible because the candidate must understand the effects of changing cool down rates on the fill rate in the pressurizer. In addition, increasing the setpoint on PCV-135 would reduce the letdown flow which would be plausible if the fill rate was reduced; however, it is not correct since fill rate will actually increase. Technical

References:

2POP-3.3

                                             -=~~~---------------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-POP002 - 1 Question Source: Bank# IPEC Bank Note changes or Modified Bank # _____ attach parent New x Question History: Last NRC Exam: -NA---- Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 _ _---->(--I.b)'-5_ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier# 2 Group # 2 KIA # 033000K402 Knowledge of Spent Fuel Pool Cooling System design feature(s) and/or interlock(s) , which provide for the following:

                                                              - Maintenance of spent fuel cleanliness Importance          2.5 Question #         59 Core offload is in progress at Unit 2. Assuming no changes are made to Spent' Fuel Pool cleanup systems, how does the core offload affect maintenance of pool clarity?

A. Increasing pool temperature leads to an increase in thermal currents. Thts causes silica material on the bottom of the pool to become suspended solids, so clarity is degraded. B. Increasing pool temperature leads to higher solubility of suspended solids, so clarity improves or remains the same. C. Increasing pool temperature leads to improved purification resin efficieooy, so clarity improves or remains the same. D. Increasing pool temperature leads to lower purification resin efficiency, so clarity is degraded. Answer: A Explanation/Justification: A. Correct. This is a particular issue at IP2 because of particles on bottom of pool for boron plates in racks Temperature goes up and causes thermal currents that stirs up debris. B. Incorrect but plausible. Solubility is affected by temperature so this is plausible however the debris that affects clarity is not soluble. C. Incorrect but plausible. Temperature affect resin efficiency, but soluble particles are not what degrades clarity. D. Incorrect but plausible. Temperature affect resin efficiency, but soluble ,. particles are not what degrades clarity.

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 2 KJA# 034000A303 Ability to monitor automatic operation of the Fuel Handling System, including: - High flux at shutdown Importance 2.9 Question # 60 The following conditions exist in the plant:

The unit is in MODE 6
All control rods are UNLATCHED
N31 reads 10 cps
N32 reads 20 cps The High Flux at Shutdown Alarms are set to:
50 cps on N31
63 cps on N32 The shift manager issues a copy of 2-SOP-13.2 (Setting the High Flux at Shutdown Alarm) and gives permission to perform the procedure. To complete and exit the procedure which of the following actions will have to be taken?

A. N31 High Flux at Shutdown setpoint must be reduced. B. N32 High Flux at Shutdown setpoint must be raised. C. Both N31 and N32 High Flux at Shutdown setpoints are within allowable range. D. Both N31 and N32 High Flux at Shutdown setpoints require adjustment. Answer: A ExplanationlJustification: The setpoint for the high flux at shutdown alarm is 1/2 decade above background. The candidate may believe that the background countrate should be multiplied by 5 vice 3.17 to calculate the setpoint.

A. Correct. The actual setpoint should be 31.7 cps. The minimum should be 25 cps and the maximum should be 38 cps. B. Incorrect. The actual setpoint should be 63.7 cps. The minimum should be 50 cps and the maximum should be 76 cps. C. Incorrect. Candidate should recognize that N31 is set too high. D. Incorrect. N32 does not require adjustment. Technical

References:

2-S0P-13.2 Proposed References to be provided: Learning Objective: 12LP-ILO-ICEXC - 11 Question Source: Bank # Note changes or Modified Bank # ___ X_ _ attach parent New Question History: Last 2 NRC Exams at IPEC: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: x 10 CFR Part 55 Content: 55.41 (b) 7 55.43 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 2 KIA # 035000K601 (\ 1~~nowledge of the effect of a

                                             ~. ~~ss or malfunction of the
                                   /'). I Of'~         fOIiOtin g will have on tj).f; t/J'( 5)rr}          S!G: - MSIVs             "{.

1.7> y 1 Importance 3.2 Question # 61 Given the following:

Unit 2 was operating at approximately 17% power during a power increase. ~
Rod Control is in manual.
24 MSIV failed closed.

Which of the following identifies the plant response approximately 10 minutes? Assume no operator acl~ "'"

                                                        ~o~iS eve~      after ,.P<

24 SG pressure Turbine Load Reactor n. PRZR Level A. Lower Tripped Tripped Lower I Approximately i i B. Higher Tripped the Same Higher I Approximately ! Approximately C. I Higher the Same

the Same Higher Approximately D. Lower the Same Lower Lower I Answer: B Explanation!Justification:

The reactor will NOT trip with a turbine trip below P-8 (18% power). With Rod, Control in Manual, rods will not step in to reduce Average Tavg. The Steam Dumps will open and remain open. Loop 24 temperature will rise to approximately Thot.

A. Incorrect. 24 SG Pressure will increase due to higher Loop Tavg in 24 loop. The turbine will trip due to the MSIV closure. The reactor will not trip. Pressurizer level will be higher due to increase in Average Tavg. B. Correct. With loop 24 at Thot, the SG pressure will increase. The turbine will trip due to the MSIV closure. The reactor will not trip. Pressurizer level will be higher due to increase in Average Tavg. C. Incorrect. 24 SG Pressure will increase due to higher Loop Tavg in 24 loop. The turbine will trip due to the MSIV closure. The reactor will not trip. Pressurizer level will be higher due to increase in Average Tavg. D. Incorrect. 24 SG Pressure will increase due to higher Loop Tavg in 24 loop. The turbine will trip due to the MSIV closure. The reactor will not trip and power will remain approximately the same on the steam dumps. Pressurizer level will be higher due to increase in Average Tavg. Technical

References:

2-E-O Proposed References to be provided: None Learning Objective: 12LP-ILO-MTG001 - 5 Question Source: Bank# IPEC Bank

                                               ----         Note changes or Modified Bank #                  attach parent New                      x Question History:                    Last NRC Exam:        NA Memory or Fundamental Question Cognitive Level:            Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _-->(........ b)'-.-5_ _ _ __ 55.43 (b) 14 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 2 KIA # 041000K302 Knowledge of the effect that a loss or malfunction of the SOS will have on the following: RCS Importance 3.8 Question # 62 The plant is operating at 100 percent power when one (1) High Pressure (HP) Condenser Steam Dump valve fails full open. Which ONE (1) of the following statements best describes the expected plant response with NO operator action? A. An Over Temperature Delta Temperature reactor trip will occur

8. A Feed Flow / Steam Flow Mismatch reactor trip will occur C. The plant will stabilize; 100 percent reactor power and less than 100 percent turbine power O. The plant will stabilize; greater than 100 percent reactor power and Tave less than programmed Tave Answer: D Explanation/Justification:

Justification: UFSAR 14.1.11, Excessive Load Increase Incident, describes an event in which "a rapid increase in the steam flow that causes a power mismatch between reactor core power and the team generator load demand." For all cases evaluated, the UFSAR states, "the plant rapidly reaches a stabilized condition at the higher power leveL" In addition, Tave will decrease slightly to add (+) reactivity to compensate for the power defect. Answer D correctly states these conditions, and is the correct choice. A. Incorrect because the OT/DT setpoint is based on not exceeding DN8R, and the UFSAR states that for a 10% step load increase, (one steam dump valve is app. 4% steam flow), the ON8R remains above the safety analysis limit DN8R value.

8. incorrect because the Feed/Steam flow mismatch trip is in coincidence with S/G low level.

C. incorrect because turbine load will not change, so the added steam demand will cause reactor power to increase. D. Corrct Technical

References:

System Description 18.0 Proposed References to be provided: None Learning Objective: 12LP-ILO-SDSHP - 9 Question Source: Bank #

                                             - -x- - IPEC Bank 5118 Note changes or Modified Bank #                  attach parent New Question History:                 Last NRC Exam:         NA Memory or Fundamental Question Cognitive Level:         Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _---->(--/.b)'--5_ _ _ _ __ 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 2 KfA# 071000A426 Ability to manually operate and!or monitor in the control room: - Authorized waste gas release, conducted in compliance with radioactive gas discharge permit Importance 3.1 Question # 63 A Gas Decay Tank release is planned. Which of the following identifies who can authorize the release at the specified limits? , Annual Average Instantaneous i Limit Quarterly Limit Limit A. CRS SM Site OM B. i CRS SM GMPO i I: RO CRS SM I RO Site OM GMPO Answer: D Explanation!Justification: SOP-5.2.4 Calculation and Recording or Radioactive Gaseous Releases Precaution and Limitation 2.6 identifies permission required for release below Annual Avg - RO, CRS, SM Quarterly Average Limit - Site Ops Manager Instantaneous Limit - General Manager Plant Operations. A. Incorrect. Plausible because the CRS can approve the Annual Limit; however the SM cannot approve the Quarterly Limit and the Site OM cannot approve the Instantaneous Limit

B. Incorrect. Plausible because the CRS can approve the Annual Limit; however the SM cannot approve the Quarterly Limit BUT the GMPO can approve the Instantaneous Limit C. Incorrect. Plausible because the RO can approve the Annual Limit; however the CRS cannot approve the Quarterly Limit and the SM cannot approve the Instantaneous Limit D. Correct Technical

References:

2-S0P-5.2.4

                                        ~~~~~-------------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-GWR001 - 5 Question Source: Bank# IPEC Bank Note changes or Modified Bank # _____ attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 13 55.43 (b) 4 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 2 KJA# 072000K101 Knowledge of the physical connections and!or cause effect relationships between the ARM system and the following systems: - Plant ventilation systems Importance 3.1 Question # 64 Given the following plant conditions:

Unit 2 is operating at 100% power.
A containment pressure relief is in progress.
A small leak develops inside containment on 342, Loop 21 Letdown Stop Valve, bonnet.

Which ONE of the following identifies the radiation monitor(s) that could have initiated the Containment Vent Isolation (CVI) signal, and the expected radiation monitor(s) response after the CVI? Note the following nomenclature: R-41 , Containment Particulate R-42, Containment Gas R-44, Plant Vent Gas & Iodine I Radiation Monitor Radiation Monitor Readings after the CVI A. R-41 OR R-42 only. R-41 , R-42 and R-44 would decrease. B. R-44 only. Only R-44 would decrease. C. R-41 OR R-42 OR R-44. R-41 , R-42 and R-44 would decrease. D. R-41 OR R-42 OR R-44. Only R-44 would decrease. Answer: D Explanation!Justification:

Any of the 3 radiations monitors, R-41/42 and R-44 will cause a Containment Ventilation Isolation. R-41/42 sample containment atmosphere and the leak is not terminated, the response on these monitors will not decrease. R-41/42 Isolate on a SI signal and indications do decrease for that condition. R-44 samples the Plant Discharge Duct. Since it is located downstream of Purge System, it will decrease after the CVI terminates the Containment Purge. A. Incorrect. Plausible because R-41/42 cause CVI, but R-44 will also cause CVI. Only R-44 indication will decrease after CVI B. Incorrect. Plausible because R-44 cause CVI, but R-41 142 will also cause CVI. Only R-44 indication will decrease after CVI C. Incorrect. Plausible because all 3 Radiation monitors cause CVI. Only R 44 indication will decrease after CVI D. Correct System Description 12.0 Technical

References:

2-S0P-12.3.3 Proposed References to be provided: None Learning Objective: 12LP-ILO-RMS001 - 3 Question Source: Bank# IPEC Bank Note changes or Modified Bank #

                                              - -X- -       attach parent New Watts Question History:                    Last NRC Exam:        Bar 2009 Memory or Fundamental Question Cognitive Level:            Knowledge:                              X Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 (b) 11 55.43 (b) 4 Comments:

Exam Outline Cross

Reference:

Level RO Tier # 2 Group # 2 KIA # 045000K523 Knowledge of the operational implications of the following concepts as they apply to the MT/G System: - Relationship between rod control and RCS boron concentration during TIG load increases Importance 2.7 Question # 65 Given the following:

Reactor power was reduced to 70% to perform a repair to one Main Boiler Feedpump.
Repairs took five days.
PRNls were adjusted to exactly match thermal power at 70%
A reactivity plan was developed to return the reactor to 100%.
The plan assumed that Control Bank 0 would be at 180 steps at the start of the power ascension. t1 A _ 0 "
Actual rod position was 200 steps on Control Bank DoOtt '~'_~~~~

How does this difference in rod position effect the amount of dilution required ana-PRNI adjustment required to return the plant to a normal 100% lineup with the PRNls properly adjusted? A. Greater dilution will be required, but less PRNI adjustment will be required. B. Greater dilution will be required, and more PRNI adjustment will be required. C. Less dilution will be required, but more PRNI adjustment will be required. D. Less dilution will be required, and less PRNI adjustment will be required. Answer: B ExplanationlJustification:

As power is raised dilution will be required and PRNls are adjusted down (if they were adjusted at lower power). The reason for the NI adjustment is that at higher temperatures more neutrons will leak and get to PR detectors. Having higher than expected boron at 75% will make the adjustment at full power greater than. otherwise. Having higher boron at 75% will require greater than normal dilutions as well. All choices are plausible because of the posibility of misunderstanding how the plant responds to a normal load increase and/or how higher rod height . will affect this (Le. higher boron) A. Incorrect B. Correct C. Incorrect D. Incorrect Technical

References:

Proposed References to be provided: None Learning Objective: 12LP-ILO-MTG001 - 5 Question Source: Bank# - - - - - IPEC Bank Note changes or Modified Bank # attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 1 ,5 55.43 (b) 6 Comments:

Exam Outline Cross

Reference:

Level Tier# Generic Equip Group # Control KIA # 1940012201 Equipment Control - Ability to perform pre-startu.12 procedur~s for the facility, incruding operating those controls associated with plant , equipment that could affect ffia~~~ I Importance 4.5 Question # 66 Plant conditions:

Plant startup in progress
Rods are 500 pcm above estimated critical position and Reactor is not critical From the list below, identify the ONE statement that describes required action(s) for this condition.

A. Maintain rods at current position and perform a re-evaluation of jnputs and mathematics used to determine ECP. B. Fully insert all control banks and perform a re-evaluation of inputs and mathematics used to determine ECP. . C. Manually trip the Reactor and perform a re-evaluation of inputs and mathematics used to determine ECP. D. Maintain rods at current position and perform a re-evaluation of inputs and mathematics used to determine ECP. Obtain Operations Manager's approval to resume approach to criticality. Answer: B Expla nationlJustification: Per POP-1.2 Att 2, the correct response is to fully insert control banks and then evaluate inputs and math of ECP. There are additional actions as well, but these do not figure in the question choices.

A. Incorrect but plausible. The direction to insert control banks is conservative. It is plausible that the procedure could have us leave rods as is while evaluation takes place. B. Correct C. Incorrect but plausible. Tripping the Reactor is more extreme than inserting control banks, but it is plausible. D. Incorrect but plausible. Ultimately the procedure will have us resume approach to criticality with OM permission, but not without first inserting controling banks. Technical

References:

2-POP-1.2 ATT. 2 Proposed References to be provided: _N-'-o..;...n_e'--__________________ Learning Objective: Question Source: Bank#

                                               ---    x--   IPEC Bank Note changes or Modified Bank #     _ _ _ _ attach parent New Question History:                    Last NRC Exam:        TMI2003 Memory or Fundamental Question Cognitive Level:            Knowledge:                                 X Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 (b) 6.10 55.43 (b) 6 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Conduct of Group # Ops KIA # 1940012142 Conduct of Operations Knowledge of new and spent fuel movement procedures. Importance 2.5 Question # 67 Which ONE of the following is the responsibility of the on-watch Reactor Operators during core re-Ioad? A. Monitor source range count rate during core reload, and remain cognizant of 11M plot results. B. Maintain continuous communications with the Refueling Floor and Outage Control Center. C. Maintain a 11M plot during fuel shuffle. D. Update the Fuel Tracking Software for each core alteration as it is performed. Answer: A ExplanationlJustification: The KIA is for conduct of operations and knowledge of refueling procedures. A Reactor Operator is used during fuel movement as the Refueling Monitor. This individual (who can actually be licensed on the other unit) is not part of the control room watch team. A. Correct B. Incorrect but plausible. An RO does this, but not the watch ROs C. Incorrect but plausible. 11M is maintained by the Refueling Monitor and Reactor Engineer, not the watch ROs D. Incorrect but plausible. This is also often done by the RO who is part of refueling group Technical

References:

EN-OP-115 Proposed References to be provided: None Learning Objective: 12LP-ILO-FHD001 - 18

Question Source: Bank# - -x- - IPEC Bank Note changes or Modified Bank # _ _ _ _ attach parent New Beaver Valley Question History: Last NRC Exam: 2005 Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 7 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Conduct of Group # Ops KIA # 1940012103 Conduct of Operations Knowledge of shift or short term relief turnover practices. Importance 3.7 Question # 68 Given the following plant conditions:

The Unit is in Mode 2 following a refueling outage.
A reactor startup is in progress lAW 2-POP-1.2, Reactor Startup.
Due to delays in the startup, the on-coming shift has arrived in the control room for shift relief.

  ~ 0 B k C is at 5 steps and cou ts                e table.

W .Ich O~E of the following~~Oj~ct CO!?rn~in. s *ft turnover I~""W Qj\P-002, Shift R~hef and Turnover?~£rI.L<} c *- , __ Wrr '( c A. Turnover can occur at any stable point (e.g., a doubling) during the start up with the approval of the Shift Manager. B. Turnover during the approach to criticality shall be avoided. The shift can be turned over when the startup is complete or reactor placed in a stable condition. C. Turnover during the approach to criticality should be avoided. The shift can be turned over ONLY with the approval of the General Manager Plant Operations. D. Turnover can occur at any stable point (e.g., a doubling) during the startup as long as NO other evolutions are in progress. . Answer:

             -B- -

Explanation!Justification:

Step 4.1.3 of OAP-002 states: Shift turnover SHALL NOT conducted during plant transients or during major steps of an evolution (i.e., Ignificant load changes, etc.). Step 4.1.9 of OAP-002 st tes: IF Reactor startup is in ~ogress. THEN watch releif in CCR SHALL NOT begin until the startup is com the Reactor is placed in a safe stable condition A. Incorrect B. Correct C. Incorrect D. Incorrect Technical

References:

OAP-002

                                           ~--------------------------

Proposed References to be provided: None Learning Objective: IOLP-ILO-ADM01 - 1 Question Source: Bank# -------- IPEC Bank Note changes or Modified Bank # ________ attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) Comments:

Exam Outline Cross

Reference:

Level Tier# Generic Equipment Group # Control KIA # 1940012207 Equipment Control - KnQwledge of the process for conducting special or infrequent tests. Importance 2.9 Question # 69 Which ONE of the following surveillance tests is required to be designated as an Infrequently Performed Test or Evolution? A. 2-PT-2M5, Safety Injection System Train A Actuation Logic and Master Relay Test B. 2-PT-SA067, Cable Spread Halon System C. 2-PT-2YOOSA, 21 EDG Mechanical Overspeed Trip D. 2-PT-Q4S, AMSAC Logic Answer: C Explanation/Justification: A. Incorrect. Plausible because this test has the potential to cause a reactor trip; however it does not meet the guidance in EN-OP-116 primarily the test is performed more frequently than quarterly and is covered by an existing approved procedure. B. Incorrect. Plausible because this test has a potential safety and equipment inoperability risk; however it does not meet the guidance in EN OP-116 primarily the test is covered by an existing approved procedure. C. Correct. EN-OP-116 states any test that actually overspeeds a turbine or Emergency Diesel Generator. D. Incorrect. Plausible because this test has the potential to cause a reactor trip; however it does not meet the guidance in EN-OP-116 primarily the test is performed more frequently than quarterly and is covered by an existing approved procedure. Technical

References:

EN-OP-116

                                         -=~~~~---------------

Proposed References to be provided: -'N_o-'n--'e_ _ _ _ _ _ _ _ _ _ __

Learning Objective: IOLP-ILO-ADM01 -1 Question Source: Bank# - - - - IPEC Bank Note changes or Modified Bank # - -x- - attach parent New North Anna Question History: Last NRC Exam: 2008 Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 _ _------>(--<b)'--5_ _ _ __ 55.43 (b) 10 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Equipment Group # Control KIA # 1940012241 Equipment Control - Ability to obtain and interpret station electrical and mechanical drawings. Importance 3.5 Question # 70 Given partial excerpt from the Chemical and Volume Control flow diagram (9321 F-2736), which of the following describes the significance of broken/dashed lines used for TCV-130? A. The valve is included here for information. The actual valve with details appears with solid lines on another drawing. B. The valve seat is undercut or valve is mechanically prevented from fully closing. There is always some minimum flow through this flowpath. C. Operation of this valve has the potential to cause a change in reactivity. D. The interfacing CCW system is classified as a "closed system" for interface LOCA. Answer: A

-~

ExplanationlJustification: A. Correct. The "ghosted" component indications are used for information and should not be used for tagging purposes because additional detail may exist on another drawing. B. Incorrect. Plausible because some valve are undercut to prevent thermal locking during heatup and cooldown activities. Also some valve have a minimum/maximum opening or closing mechanical stop. C. Incorrect. Plausible because TCV-130 has the potential to affect reactivity by causing letdown excessive cooling or heatup.

D. Incorrect. Plausible because the FSAR classifies some systems as Open or Closed. Isolation valves for all fluid system lines penetrating the containment provide at least two barriers against leakage of radioactive fluids to the environment in the event of a loss-of-coolant accident. These barriers, in the form of isolation valves or closed systems, are defined on an individual line basis. Technical

References:

Proposed References to be provided: _9_3_2_1_-2_7_3_6....,,(. P_a_rt_ia.

                                                                      ...       .I). . _ _ _ _ _ __

Learning Objective: Question Source: Bank# IPEC Bank Note changes or Modified Bank # _ _ _ _ attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 _ _---->(..c...Jb)'----3_ _ _ __ 55.43 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Rad Group # Controls KIA # 1940012305 Radiological Controls - Ability to use radiation monitoring systems, such as fixesd radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc. Importance 2.9 Question # 71 Given the following conditions:

The plant is at 100% power
23 Large Gas Decay Tank is aligned for in-service and re-use
24 Large Gas Decay Tank is in standby
22 Large Gas Decay Tank is isolated with a pressure of 90 psig and a content of 5000 Curies
All remaining Gas Decay Tanks are inerted with nitrogen
22 Large Gas Decay Tank relief valve (1622) fails open
No radiation monitors were in alarm prior to the 1622 failure Which ONE of the following describes the plant response to this event?

A. High radiation level alarm on R-50, Waste Gas Decay Tank Monitor AND ,n1.__ R-44, Plant Vent Air Monitor. B. High radiation level alarm on R-44, Plant Vent Air Monitor. R-50, Waste Gas Decay Tank Monitor does NOT alarm. C. High radiation level alarm on R-50, Waste Gas Decay Tank Monitor. R-44 , Plant Vent Air Monitor does NOT alarm. D. NO high radiation level alarm on R-50, Waste Gas Decay Tank Monitor OR R-44 , Plant Vent Air Monitor. Answer: B ExplanationlJustification:

For this situation, the tank will relieve directly to the plant vent and be monitored by R-44. R-50 should be unaffected by this leak. A. Incorrect but plausible. Plausible because a candidate may wrongly assume that R-50 would go up B. Correct C. Incorrect but plausible. Plausible because a candidate may believe R-50 would go up and may have a misconception of where this relieves to (e.g. WHUT) or that R-44 will not alarm. D. Incorrect but plausible. Plausible because a candidate may believe neither monitor will alarm or may think it relieves to a closed tank. Technical

References:

System Description 12.0 Proposed References to be provided: None Learning Objective: 12LP-ILO-GWR01 - 13 Question Source: Bank # ---- X - - - IPEC Bank 8370 Note changes or Modified Bank # ________ attach parent New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 12, 13 55.43 (b) 4 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Rad Group # Controls KIA # 1940012312 Radiological Controls Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filte-ers,-,-,-,-e_tc_._ _ _. Importance 3.2 Question # 72 Unit 2 is at 1% Reactor power coming out of a refueling outage. Personnel are in containment making adjustments to 23 RCP vibration probes. The CRS and SM decide they want to raise power to 2% in preparation for power ascension later that day. Based on OAP-007, Containment Entry and Egress, what is required regarding this power ascension? A. Personnel working on the RCP vibration probes will have to move to the outer crane wall. When the power increase is complete the workers can return to the RCP. B. Power can be raised. Since the plant will remain in a mode below Mode 1 dose rate changes will be minimal, so the power ascension does not require additional action per OAP-007. C. Power can be raised. However, since there are personnel in the inner crane wall, OAP-007 requires the SM to specifically approve the power ascension. D. The RP Supervisor and entry party must be notified prior to any planned change in power level. The RP Supervisor will then decide if workers need t)vr('~ to-lilEh'C prior to raising power. {/Vlolf'/.I'

~~

Answer: 0 ExplanationlJustification: This situation actually occurred at IP3, which led to the procedural requirement. A. Incorrect but plausible. It is not unreasonable that OAP-OOl would have required removing personnel prior to power ascension, not just moving to outer crane wall. B. Incorrect but plausible. It would be reasonable to assume that this power change would have minimal effect on dose rates, but this is not true. C. Incorrect but plausible. The SM is often allowed to authorize items that require slightly greater levels of control and decision making. Based on B above discussion, it is reasonable that a candidate may assume this change will have minimal effect. D. Correct based OAP-OOl step 2.22 Technical

References:

OAP-OOl Proposed References to be provided: _N-'o_n_e_ _ _ _ _ _ _ _ _ _ __ Learning Objective: IOLP-ILO-ADM01 - 4 Question Source: Bank#

                                                       - - - IPEC    Bank Note changes or Modified Bank #                   attach parent New                         x Question History:                       Last NRC Exam:        NA Memory or Fundamental Question Cognitive Level:               Knowledge:                             x Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5, 6 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Rad Group # Controls KIA # 1940012307 Radiological Controls - Ability to comply with radiation work permit requirements during normal and abnormal conditions. Importance 3.5 Question # 73 A Radiation Work Permit (RWP) is needed to enter a very high radiation level area during a backshift to stop a minor leak. No additional personnel or equipment hazards exist. Who is required to authorize the RWP? A. AnySRO B. The SM C. The Watch RP Technician D. RP Supervisor Answer: D Explanation!Justification: A. Incorrect but plausible B. Incorrect but plausible C. Incorrect but plausible D. Correct as per EN-RP-105 Section 4 item 2 Technical

References:

EN-RP-105

                                         -=~~~~------------------

Proposed References to be provided: _N_o_n_e_ ________________ Learning Objective: IOLP-ILO-ADM01 - 1 Question Source: Bank# IPEC Bank

Note changes or Modified Bank # _ _X_ _ attach parent New DC Cook Question History: Last NRC Exam: 2001 Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 4 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Emerg Group # Proc/Plan KIA # 1940012419 Emergency Procedures/Plan Knowledge of EOP layout,

                                                  ~mbols, and icons.

Importance 3.4 Question # 74 Given the following conditions:

The crew is responding to a large break LOCA
A CORE COOLING status tree ORANGE path causes a transition to FR-C.2, Response to Degraded Core Cooling
During performance of FR-C.2, the CORE COOLING status tree changes from ORANGE to YELLOW
An ORANGE path exists on the CONTAINMENT status tree Which ONE of the following describes the required action(s)?

A. Complete FR-C.2 and then go to FR-Z.1 , because a functional restoration procedure must be completed unless preempted by a higher priority condition. B. Go to FR-Z.1, because an ORANGE path has higher priority than a YELLOW path. Completion of FR-C.2 is not needed. C. Go to FR-Z.1, then complete FR-C.2 because the CORE COOLING status tree had been in an ORANGE path. D. Perform FR-C.2 and FR-Z.1 concurrently, because FR procedures of the same priority can be executed together. Answer: A Explanation/Justification: A. Correct Answer: Step 4 ..3.13 of OAP 12 requires the completion of a FRP entered due to a RED or ORANGE condition unless that FRP is preempted by a higher priority condition. B. Orange is higher priority than Yellow, but OAP 12 step 4.3.13 requires the completion of the current procedure.

C. FR-C.2 has higher priority than FR-Z.1 and needs to be completed first in accordance with OAP 12 step 4.3.13. D. FR-C.2 is the higher priority and needs to be completed first in accordance with OAP 12 step 4.3.13. Technical

References:

OAP-12 Proposed References to be provided:

                                         . .:. .N. .;. .o. :. . :n.:. .;.e=---_________________

Learning Objective: 12LP-ILO-EOPROU - 12 Question Source: Bank#

                                                               ------          x    IPEC Bank Note changes or Modified Bank #                        ______ attach parent New Question History:                   Last NRC Exam:                                  NA Memory or Fundamental Question Cognitive Level:           Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Emerg Group # Proc/Plan KIA # 1940012429 Emergency Procedures/Plan Knowledge of the emergency plan. Importance 3.1 Question # 75 Given the following:

A Site Area Emergency has been declared.
The Emergency Response Organization is staffed.
A repair team consisting of 1 NPO, 1 mechanic, and 1 HP technician must be sent to the PAS to isolate a leak.

Which ONE of the following Emergency Response Facilities is responsible for assembly and preparation of the team? A. Control Room B. Technical Support Center (TSC) C. Operational Support Center (OSC) O. Emergency Operations Facility (EOF) Answer: C Explanation/Justification: A. Incorrect. Make initial declaration and classification of event. Manipulation of the reactor or plant to mitigate the consequences of an accident remain the primary function of the CR. Plausible because before Emergency Response facilities are manned, the CR would direct this action. S. Incorrect. The TSC is the central facility for the accumulation and re transmittal of plant parameters. The TSC provides Technical Support. Plausible because candidate may confuse functions performed by different facilities.

C. Correct. The OSC is where survey, operations and repair teams are dispatched into areas of the plant and is the staging area for individual who may be assigned. D. Incorrect. The EOF provides overall management of the Indian Point response. Technical

References:

IP-EP-230 Proposed References to be provided: ----------------------------- _N-'-o.:...;n-'-e:.-...._________________ Learning Objective: IOLP-ILO-ERT001 - 1 Question Source: Bank# IPEC Bank

                                                     ----            Note changes or Modified Bank #                         x   attach parent New Ginna Question History:                  Last NRC Exam:                    2007 Memory or Fundamental Question Cognitive Level:          Knowledge:                                         x Comprehension or Analysis:

10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level SRO Tier # 1 Group # 1 KIA # 000008A213 Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: - High pressure safety injection pump flow indicator, ammeter, and controller Importance 3.9 Question # 76 Given the following:

A Pressurizer Safety Valve failed open.
23 SIP is out of service for maintenance.
All other equipment functions as designed.
RCS Temperature stabilized at approximately 530°F.
RCS Pressure stabilized at approximately 950 psig.
Approximately 15 minutes after the safety injection actuation 21 SIP tripped on overcurrent.
The team has just transitioned to E-1 Loss of Reactor or Secondary Coolant.
E-O, Reactor Trip or Safety Injection, Attachment 1 is in progress.

Which of the following correctly states the expected SI flowrate indications and procedural actions for this condition? A. Approximately 0 gpm to each RCS loop. Establish SI flow using E-O Attachment 1 . B. Approximately 0 gpm to each RCS loop. Do not use E-O Attachment 1 to establish flow since E-1 has been entered and will address this condition. C. Approximately 100 gpm to each RCS loop. The EOPs will not require any adjustments to SI flow. D. Approximately 200 gpm each to loops 22 & 24. E-1 will allow for balancing SI flow if desired.

Answer: A ExplanationlJustification: 21 SIP supplies loops 21 & 23. 23 SIP supplies loops 22 & 24. 22 SIP supplies all loops unless 21 or 23 pump breakers are open with an SI signal present. If 21 SIP breaker is open with an SI signal present, 851B (discharge valve for 22 SIP to loops 22 & 24) will automatically close. If 23 SIP breaker is open with an SI signal present, 851A (discharge valve for 22 SIP to loops 21 & 23) will automatically close. This condition is only addressed in E-O Attachment 1 A. Correct B. Incorrect. Plausible because the flowrate is correct, but the condition is not addressed by E-1. C. Incorrect. Plausible because the flowrate is correct for22 SIP; however, with both 21 & 23 SIPs tripped, 851 AlB will automatically close. Candidate may believe that 22 SIP is supplying all loops. D. Incorrect. Plausible because the flowrate is correct for 22 SIP; however, with both 21 & 23 SIPs tripped, 851 AlB will automatically close. Candidate may believe 851B will remain open. In addition, E-1 will not address this condition. Technical

References:

E-O Attachment 1

                                          ~----~~-------------------

Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPEOO - 3~_ _ _ __ Question Source: Bank# IPEC Bank Note changes or Modified Bank # ______ attach parent New x- - - - - - Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _----3(....:....b)t.-_ _ __ 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level SRO Tier # 1 Group # 1 KIA # 0000112408 Emergency Procedures/Plan Knowledge of how abnormal operating procedures are used in conjunction with EOPs. --------, Importance 4.5 Question # 77 Given the following conditions:

A loss of Instn.lment Air has occurred.
The CRS has directed a reactor trip in accordance with the requirements of AOP-AIR-1, Air System Malfunctions.
When the Reactor was tripped, a Large Break LOCA occurred.
E-O, Reactor Trip or Safety Injection has just been entered.

Which ONE of the following describes the allowable usage of AOP-AIR-1 while responding to this event? ? A. Whe~~mediate actions are complet~, resume AOP-AIR-1 until all actionS- are completed. Verification of automatic actions in E-O cannot be performed with a loss of Instrument Air. B. Discontinue use of AOP-AIR-1 until transition to any recovery procedure. Parallel use is only allowed when E-O is complete.

             £,-0 C. Whe'1 immediate actions are complete, parallel use of AOP-AIR-1 is allo\Ak'd when performance will not detract from performance of EOPs.

D. Discontinue use of-AOP-AIR-1. The EOP network will direct actions to restore Instrument Air to vital components. Answer: C Explanation/Justification:

From OAP-015 AOP Users Guide 4.1.18 IF directed to INITIATE actions in a referenced procedure OR attachment THEN the actions should be taken while continuing on in the AOP. 4.1.18.1 IF an AOP directs the initiation of E-O, THEN the AOP actions will normally be taken after transition to ES-0.1 (Reactor Trip Response) or after step 4 of E-O, (Reactor Trip Or Safety Injection) WHEN performance will NOT detract from performance of the EOP. 4.1.18.2 The CRS may delegate the completion of AOP actions while continuing in the EOPs. A. Incorrect. Plausible because the actions of AOP-AIR-1 may restore air pressure and make verification of auto actions proceed more smoothly. B. Incorrect. As shown above, an AOP can be performed in parallel with an EOP. Plausible because the OAP does state that the AOP will NORMALLY be resumed when transition to ES-0.1. C. Correct D. Incorrect. Plausible because some AOPs direct an unconditional exit to E-O. For those AOPs, no further actions are taken in the AOP. Technical

References:

OAP-015 Proposed References to be provided: ~-------------------------- _N __o__n__e '--________________ Learning Objective: 12LP-ILO-EOPROU - 19 Question Source: Bank# IPEC Bank Note changes or Modified Bank # __X _ _ attach parent 16887 New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level SRO Tier# 1 Group # 1 KIA # 000011A213 Ability to determine and interpret the following as they apply to a Large Break LOCA:

                                                  - Difference between overcooling and LOCA indications Importance                              3.7 Question #     78 Given the following conditions:

The reactor has tripped. Safety Injection and Containment Spray have actuated.
The team is performing the actions of E-1, Loss of Reactor or Secondary Coolant
A Red Path exists on the Integrity Status Tree
The CRS directs transition to FR-P.1, Response to Imminent Pressurized Thermal Shock
The procedure immediately sends the team back to E-1 Which of the following identifies the plant parameters checked to immediately exit FR-P.1 and why FR-P.1 is not implemented?

A. RCS Pressure and SG Pressure. SG pressure greater than RCS pressure indicates a Large Break LOCA vice a Steam Break; the excessive cooldown will not continue. B. RCS Pressure and RHR Flow. RHR Flow greater than the minimum value inidicates a Large Break LOCA and thermal shock is not a serious concern for this event. C. Containment Radiation and RHR Flow. Elevated Radiation and RHR flow above the minimum value indicate a Large Break LOCA; repressurization of the RCS is virtually impossible during a Large Break LOCA. D. RCS Pressure and Containment Pressure. RCS Pressure and Containment Pressure approximately equal indicate a Large Break LOCA. The actions in FR-P.1 will delay the actions in ES-1.3, Transfer to Cold Leg Recirculation, causing a potential loss of core cooling.

Answer: B ExplanationlJustification: FR-P.1 step 1 checks RCS pressure and RHR Flow to identify a Large Break LOCA. A. Incorrect. Plausible because RCS pressure will be less than SG pressure on a large break LOCA; however, the cooldown will continue until transfer to recirc. B. Correct. From the background document: For transients where RCS pressure is less than the RHR pump shutoff head and flow from the RHR pumps has been verified, the operator should return to the procedure and step in effect since these symptoms are indicative of a large-break LOCA. In this instance, the actions of 2-FR-P.1 should not be performed since pressurized thermal shock is not a serious concern for a large-break LOCA. C. Incorrect. Plausible because containment radiation is the key parameter used to distinguish between a LOCA and Steam Break accident inside containment; however these are not the parameters used to identify a LBLOCA. Also, "repressurization during a LBLOCA is virtually impossible" is true. D. Incorrect. Plausible because RCS Pressure and Containment Pressure will be approximately equal during a LBLOCA. These are not the parameters used. Also actions in ES-1.3 are time critical to establish flow before the RWST empties. Technical

References:

2-FR-P.1 Background Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPFP1 - 1 Question Source: Bank# IPEC Bank Note changes or Modified Bank # attach parent New x Question History: Last NRC Exam: Memory or Funda Question Cognitive Level: Knowledge: Comprehension or Analysis: x 10 CFR Part 55 Content: 55.41 (b) 10

Exam Outline Cross

Reference:

Level Tier # 1 Group # 1 KIA # 000040A202 Ability to determine and interpret the following as they apply to the Steam Line Rupture: - Conditions requiring a reactor trip Importance 4.7 Question # 79 Given the following plant conditions on Unit 2:

Containment Pressure: 1 psig and rising.
RCS pressure: 2225 PSIG and lowering.
Reactor power: 63% and rising.
Average Tavg: 557°F and lowering.
Turbine power: 561 MWe and lowering.

Based on the above plant indications, what event is occurring and what are the required actions/procedures to address the event? A. A Steamline Break. Trip the reactor trip, Close MSIVs and go to E-O, Reactor Trip or Safety Injection. B. A Steamline Leak. Perform a rapid Load reduction per AOP-RLR-1, Rapid Load Reduction. C. A Small Break RCS LOCA. Trip the reactor and go to E-O, Reactor Trip or Safety Injection. D. An RCS Leak. Perform AOP-LEAK-1, Sudden Increase in Reactor Coolant Leakage. Answer: A Explanation/Justification: Immediate Actions of AOP-UC-1 require a Reactor Trip, Close MSIVs and go to E-O for unisolable Steam Leak.

A. CORRECT. Reactor power is rising, indicating positive reactivity event. Electric load is lowering, indicating loss of steam to the turbine. Turbine power should be closer to 670 Mw with Tavg closer to 558°F based on this reactor power. Based on this degree of mismatch and unisolable (inside VC) reactor trip is required. B. INCORRECT. A Steam Leak is occuring, but based on the large mismatch a reactor trip is required. C. INCORRECT. Plausible since RCS pressure is lowering, but Reactor power is rising - indicating positive reactivity event. D. INCORRECT. Plausible since RCS pressure is lowering, but Reactor power is rising - indicating positive reactivity event. Technical

References:

2-AOP-UC-1

                                           ~~-~------------

Proposed References to be provided: _N_o_n_e_ _ _ _ _ _ _ _ _ _ ____ Learning Objective: 12LP-ILO-AOPUC1 -1 Question Source: Bank# IPEC Bank Note changes or Modified Bank # _ _X___ attach parent New DC Cook Question History: Last NRC Exam: 2007 Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: x 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level SRO Tier # 1 Group # 1 KfA# 0000582446 Emergency Procedures/Plan Ability to verify th~ are consisttmrWffhthe plant conditions. Importance 4.2 Question # 80 Given the following:

A Safety Injection occurs while Unit 2 is operating at 100% power.
During the bus transfer, the Station Auxiliary Transformer trips and the transient results in the breaker connecting 24 Battery to 24 DC PP tripping open.
There is no fault on 24 DC Bus. All other equipment operates as designed.

Assuming no operator action to re-close the breaker, how does this affect operator response to the event? A. RWST low level alarms will not function. RWST level will have to be monitored to ensure transition to ES-1.3 Cold Leg Recirculation is not required. B. 23 EDG will start and power Bus 6A, however safeguards loads will have to be manually started in accordance with ES-O.1 Reactor Trip Response. C. Safeguards loads will automatically sequence on, but only after 23 EDG output breaker is manually closed in accordance with the Alarm Response Procedure. D. Since there is no fault on 24 DC Bus and all auto transfers of DC power occurred, response will be unaffected. Answer: A Explanation/Justification: This question comes down to understanding that alarms will not be available for this fault. 6A switchgear and 23 EDG control power will automatically swap to backup DC, but the alarm power will not. It also may not be clear to a candidate that the battery charger can not supply the bus.

A. Correct. These alarms will not function. B. Incorrect but plausible. Plausible because a candidate may not realize just what will automatically back up. C. Incorrect but plausible. Plausible because a candidate may not realize just what will automatically back up. D. Incorrect but plausible. Plausible because a candidate may not realize just what will automatically back up. Technical

References:

2-AOP-DC-1 Attachment 12 Proposed References to be provided: None Learning Objective: 12LP-ILO-EDS03 - 11 Question Source: Bank# IPEC Bank

                                              -----       Note changes or Modified Bank #                attach parent New                     x Question History:                   Last NRC Exam:        NA Memory or Fundamental Question Cognitive Level:           Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _---"(-Lb)_7_ _ _ __ 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level SRO Tier # 1 Group # 1 KIA # 0000092125 Conduct of Operations - Ability to interpret reference materials such as graphs, curves, tables etc. Importance 4.2 Question # 81 Given the following conditions:

A reactor trip with SI occurred at 1240 hours
At 1550 hours, the control room operators transitioned from E-1, Loss of Reactor or Secondary Coolant, to ECA-1.1, Loss of Emergency Coolant Recirculation
At 1600 hours, the operators are determining if SI flow can be terminated The following conditions are observed:
RWST level is 15 feet
RCS Wide Range Pressure is 1600 psig
RCS subcooling based on core exit TCs is 70°F
CNMT Pressure is 8 psig
All RCPs have been secured
RVLlS level is 69% on Natural Circulation Range Using attached procedure, which ONE of the following actions is required?

A. Terminate Safety Injection. B. Establish 235 gpm Safety Injection flow. C. Establish 275 gpm Safety Injection now. D. Establish 460 gpm Safety Injection flow. Answer: B Explanation!Justification:

The subcooling requirement for terminating SI is not met, so the ECA-1.1 table is used to determine flow. 200 minutes have elapsed, so 235 gpm required. A. Incorrect but plausible. 70 degrees meets thd non-adverse subcooling requirement to terminate SI, so this is plausible. B. Correct per ECA-1.1 step 14 and figure ECA11-1 C. Incorrect but plausible. Plausible since table can be read incorrectly. D. Incorrect but plausible. Plausible since table can be read incorrectly. Technical

References:

2-ECA-1.1 Proposed References to be provided: 2-ECA-1.1 PG 11 & 31 Lea¢lnQ Objective: 12LP-ILO-EOPC11 - 1 12LP-ILO-EOPC11 - 2 Question Source: Bank# x

                                               - - - - - - IPEC Bank - 8307 Note changes or Modified Bank #                  attach parent New Question History:                   Last NRC Exam:         NA
                                                         ~~--

Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _--->(--'b):.-_ _ _ _, 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level SRO Tier# 1 Group # 2 KIA # 0000032222 I Equipment Control Knowledge of limiting conditions for operations and safety limits. Importance 4.7 Question # 82 Reactor power is 80% during a power ascension following a forced outage. During rod motion the stationary gripper fuse blows for Rod C-3. Repairs are expected to take several days. Which of the following statements is correct regarding continued operation with this condition? A. Power can be held at the current level provided a flux map is performed within 12 hours to ensure core hot channel factor limits are not exceeded. B. Power can be held at the current level provided a flux map is performed within 12 hours to ensure core hot channel factor limits are not exceeded and safety analyses are re-evaluated within 5 days to ensure they are valid for current conditions. C. A power reduction will be required to ensure that axial flux difference limit assumptions are valid. D. A power reduction will be required to ensure core hot channel factor limits are not exceeded. Answer: -D- - Explanation/Justification: A. incorrect but plausible. 85% is significant power level for determining how far can mis-aligned, so an operator could assume 80% is low enough to allowed continued operation. There are ~2 hour T.S. requirements and a flux map is required which add to the plausibility of continued operation. B. incorrect but plausible. See above, plus there are requirements for performing this analyses.

C. incorrect but plausible. A power reduction is required, but the the reason is not related to AFD. For this dropped rod (near a single NI) AFD is only affected in one channel. D. Correct per T.S. and basis for LCO 3.1.4 Technical

References:

Tech Spe_c_s_3_.1_._4_________~__..__ Proposed References to be provided: None Learning Objective: 12LP-ILO-ICROD - 14 Question Source: Bank# ----~ x IPEC Bank 24193 Note changes or Modified Bank # ______ attach parent New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _---->(--<b)'--_ _ __ 55.43 (b) 2,5 Comments:

Exam Outline Cross

Reference:

Level Tier# Group # 2 KIA # 000024A202 Ability to determine and interpret the following as they apply to the Emergency Boration: - When use of manual boration valve is needed Importance 4.4 Question # 83 Unit 2 was operating at 100% power with no equipment out of service when the following occurred:

A loss of instrument air pressure occurred in the Primary Auxiliary Building.
Subsequently 112C (VCT Outlet Valve) closed and cannot be re-opened.
No other equipment failures occurred.

How do these failures affect operability of the Boration System specified in the Technical Requirements Manual for use in Emergency Boration of the Reactor? A. The TRO is satisfied because Boration is still available from the RWST and the boric acid storage system. Emergency Boration could be performed using MOV-333 (Emergency Boration Valve). B. The TRO is satisfied because Boration is still available from the RWST and the boric acid storage system. Emergency Boration could be performed using LCV-112B (RWST Emergency M/U Valve) which has failed open due to the loss of instrument air. C. The TRO is not satisfied because Boration is only available from the RWST. Emergency Boration could be performed using LCV-112B (RWST Emergency M/U Valve) which has failed open due to the loss of instrument air. D. The TRO is not satisfied because Boration is only available from the RWST. Emergency Boration could be performed using LCV-112B (RWST Emergency M/U Valve) which can be operated locally.

Answer: A Explanation/Justification: A. Correct B. Incorrect because LCV-112B does not fail open C. Incorrect because the TRO is satisfied and because LCV-112B does not fail open D. Incorrect because the TRO is satisfied. Technical

References:

TRM 3.1.B.1, System Description 3.0 Proposed References to be provided: _N_o"--n-'-Ce'--____________ Learning Objective: 12LP-ILO-CVC02 - 5 12LP-ILO-CVC02 -14 Question Source: Bank# IPEC Bank

                                             ----       Note changes or Modified Bank #               attach parent New                       x Question History:                  Last NRC Exam:       NA Memory or Fundamental Question Cognitive Level:          Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 2,5 Comments:

Exam Outline Cross

Reference:

Level Tier # Group # 2 KIA # 000037A202 Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: Agreement/disagreement among redundant radiation monitors Importance 3.9 Question # 84 Given the following:

A startup is in progress at 20% power.
Radiation Monitor 45 (Air Ejector Exhaust) is OOS.
Radiation Monitor 55B (22 SG Blowdown) is at Warn setpoint.
No other radiation monitor alarms exist.

Which of the following is true regarding these conditions? 7 A. Radiation Monitor 55S (22 SG Slowdown) has failed hi1(an~ should be declared inoperable. Radiation Monitor 61B (22 SG N- 6) Is'the most sensitive to SG tube leakage and should indicate a Warn condition before R-55B (22 SG Blowdown). B. Radiation Monitor 55B (22 SG Blowdown) has failed high and should be declared inoperable. Radiation Monitor 29 (22 Main Steam Line) should also be at the Warn setpoint if an actual tube leak existed. C. Radiation Monitor 55B (22 SG Blowdown) may indicate a tube leak. 2 AOP-SG-1 SG Tube Leakage will use Radiation Monitor 61B (22 SG N

16) to confirm or eliminate the existence of tube leakage.

D. Radiation Monitor 55B (22 SG Blowdown) may indicate a tube leak. The lack of redundant radiation monitors does not eliminate a tube leak. Answer: D Explanation/Justification:

the existence of a SG tu e leak. The Warn setpoint for Radiation Monitor 45 is set for an equivalent 30 pd (gallons per day) leakrate. The alarm setpoint for Radiation Monitors 61A-D is set for 5 gpd. However, below 30% power R-61A-D may not indicate accurately. A. Incorrect: Radiation Monitors 61A-D may not be accurate below 30% power per AOP-SG-1 Background Document. B. Incorrect: The setpoint for Radiation Monitor 29 may not be accurate and thus cannot be used to eliminate a SG tube leak. C. Incorrect: Radiation Monitors 61A-D may not be accurate below 30% power per AOP-SG-1 Background Document. D. Correct Technical

References:

System Description 12.0 Proposed References to be provided: None Learning Objective: 12LP-ILO-RMS001 - 5 Question Source: Bank# IPEC Bank

                                              ----         Note changes or Modified Bank #    ----         attach  parent New                     x Question History:                    Last NRC Exam:       NA
                                                        ~~--

Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 11 55.43 (b) 4 Comments:

Exam Outline Cross

Reference:

Level SRO Tier # 1 Group # 2 KIA # OOWE13234 Radiological Controls Knowledge of radiation exposure limits under normal and emergency conditions. Importance 3.7 Question # 85 Given the following:

A severe accident has occurred at Unit 2. 22 SG has extremely high activity levels due to the accident.
An over pressure condition exists on 22 SG that threatens to lift a safety valve.
Local operation of 22 SG Atmospheric Dump Valve (ADV) is required to lower SG pressure in a controlled manner.
The dose rate at the local ADV controls is 20 Rem!hr and the evolution is expected to take 30-45 minutes.
A Reactor Operator who has not entered the RCA this year has volunteered to perform the task.

Which of the following statements is correct regarding this evolution? A. The evolution cannot be performed because the individual is expected to exceed their NRC occupational limit. B. The evolution cannot be performed because individuals are limited to 10 Rem to protect valuable equipment. C. The evolution can be performed because individuals are limited to 25 Rem to save equipment necessary to protect the health and safety of the public. D. The evolution can be performed because there is no limit to save equipment necessary to protect the health and safety of the public. Answer: C Explanation!Justification:

A. Incorrect but plausible because the NRC normal limit will be exceeded. B. Incorrect but plausible. 10 Rem is the limit for saving valuable equipment, so it is a plausible value C. Correct D. Incorrect but plausible. This could be selected if the candidate fails to accurately calculate total dose or does not know NRC limit of 5 Rem/year Technical

References:

EP Form 6

                                          ~-------------------------

Proposed References to be provided: None Learning Objective: IOLP-ILO-ERT003 - 3 Question Source: Bank# IPEC Bank

                                               ~---

Note changes or Modified Bank # attach parent New x Question History: Last NRC Exam: NA

                                                          -~~-

Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: x 10 CFR Part 55 Content: 55.41 (b) 55.43 (b) 7 Comments:

Exam Outline Cross

Reference:

Level SRO Tier# 2 Group # 1 KIA # 004000A213 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Low RWST Importance 3.9 Question # 86 A large break loss of coolant accident (LBLOCA) occurs. All equipment is available at the start of the event and functions as designed. In responding to this event, which of the following pumps could be secured first, and what is the procedural guidance for this action? A. 21 RHR Pump in E-O, Reactor Trip or Safety Injection. B. 22 Charging Pump prior to manipulating Recirc Switches in ES-1.3, Transfer to Cold Leg Recirculation. C. 22 SI Pump from Recirc Switch 1 in ES-1.3, Transfer to Cold Leg Recirculation. D. 21 and 22 RHR Pumps from Recirc Switch 3 Switches in ES-1.3, Transfer to Cold Leg Recirculation. Answer: B Explanation/Justification: A. Incorrect. Plausible because for most accident conditions (not including LBLOCA) an RHR pump is secured first to prevent "Strong Pump - Weak Pump" interaction. B. Correct. This action is performed to reduce loads on the 480V buses prior to transferring to recirculation.

C. Incorrect. Plausible because 22 SIP is secured first when Recirc Switch 1 is placed to ON; however, this action is performed after the charging pump is secured. D. Incorrect. Plausible because 21 and 22 RHR Pumps are secured using Recirc Switch 3; however, this action is performed acter the charging pump is secured. Note: Recirc Switch 1 and 3 are placed to on in the same step. Technical

References:

2-ES-1.3 Proposed References to be provided: ----------------------------- None Learning Objective: 12LP-ILO-EOPS13 - 4 Question Source: Bank # IPEC Bank

                                              -----       Note changes or Modified Bank #     _____ attach parent New                      x Question History:                  Last NRC Exam:         NA Memory or Fundamental Question Cognitive Level:          Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _----->(--'b)'--_ _ __ 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level SRO Tier # 2 Group # 1 KIA # 0080002237 Equipment Control - Ability to determine operability and/or availability of safety related equipment. Importance 4.6 Question # 87 Unit 2 is cooling down in Mode 4 preparing for a refueling outage.

22 CCW was tagged out at the time of shutdown.
23 CCW trips while preparing to place RHR in service.

Which of the following is correct regarding the Tech Spec requirements for these conditions? A. Two trains of CCW are inoperable. Since there is no AOT for this condition, LCO 3.0.3 applies. B. One train of CCW is inoperable. Enter 72 hour AOT for this condition and continue cooldown. C. Mode 5 cannot be entered because two trains of RHR will be required when Steam Generators can no longer be credited for RCS heat removal. D. Only one CCW pump is required to accommodate normal and accident cooling loads, a Safety Function Determination can be performed to satisfy the CCW LCO. Answer: B Explanation/Justification: A. Incorrect but plausible. One train of CCW is still operable based on these conditions. A candidate may not realize that only one train is inoperable. B. Correct. This is not a "direct lookup" type of situation because with two inoperable pumps it is not a clear call, but a well prepared ca~'date should figure this out.

                                                                            ---     , I(

C. Incorrect but plausible. RHR cooling will be effected by this degraded CCW availability, so this answer is plausible. This is incorrect for a number of reasons. One reason is that the actual T.S. statement the plant is in requires going to Mode 5. D. Incorrect but plausible. This is incorrect since there is no way to get relief from the CCW LCO via a safety function determination. This answer is plausible because the beginning of the statement is what T.S. basis says forCCW. Technical

References:

Proposed References to be provided: Learning Objective: Question Source: Bank# IPEC Bank Note changes or Modified Bank # attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (,-b..L-)4_ _ _ __ 55.43 (b) 2 Comments:

Exam Outline Cross

Reference:

Level SRO Tier # 2 Group # 1 KIA # 0120002450 Emergency Procedures/Plan Ability to verify system alarm setpoints and operate controls identified in the alarm response manual. Importance 4.0 Question # 88 Given the following:

The reactor is at 8% power preparing to synchronize the Main Generator to the grid.
Bus 5 normal feed breaker tripped on overcurrent.
The reactor remains critical.
21 & 24 RCPs trip on under voltage.
22 & 23 RCPs are operating.

Which of the following correctly describes the plant status and what if any actions should be taken? A. No Reactor Protection Setpoints have been exceeded. Per the ARP, trip the Reactor and go to E-O, Reactor Trip or Safety Injection. B. No Reactor Protection Setpoints have been exceeded. Per 2-AOP-138KV 1 Loss of Power to 6.9KV Bus 5 and/or 6, trip the Reactor and go to E-O, Reactor Trip or Safety Injection. C. The reactor should have tripped on loss of flow in 2 loops. Per the ARP, trip the Reactor and go to E-O, Reactor Trip or Safety Injection. D. Under frequency on 2 of 4 buses should have caused all RCPs to trip. Per 2-AOP-138KV-1 Loss of Power to 6.9KV Bus 5 and/or 6, trip the Reactor and go to E-O, Reactor Trip or Safety Injection. Answer: A Explanation/Justification:

A. Correct. 2 ARP-SAF directs a reactor trip if any pump is tripped regardless of power. B. Incorrect because wrong procedure is given. PlclUsible it is reasonable that the AOP could specify tripping reactor-- ,_. C. Incorrect. Plausible because the candidate may forget that the low flow trips (and alarms) are bypassed below P-7 (10% power). D. Incorrect. Plausible because under frequency should trip all 4 RCPs; however, if power is lost, the under frequency will not trip the remaining RCPs. Technical

References:

2-ARP-SAF Proposed References to be provided: Learning Objective: 12LP*ILO-RCSRCP - 15 Question Source: Bank# IPEC Bank Note changes or Modified Bank # _____ attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _----'-{b~}_ _ _ _ _. 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level SRO Tier# 2 Group # 1 KJA# 059000A203 Ability to (a) predict the impacts of the following malfunctions or operations on the MFW System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: - Overfeeding event Importance 3.1 Question # 89 Given the following:

Unit 2 is operating at 100% when 21 MBFP trips to due to a Lovejoy malfunction.
The automatic runback functions to lower turbine load, and SG NR Levels all lower to approximately 15% due to shrink.

Which of the following describes expected SG NR response and how this is addressed 2-AOP-FW-1, Loss of Feedwater? A. Increase until a Turbine Trip occurs due to integral error in the feedwater control system. To prevent this, the AOP provides guidance to remove the integral error from the FRVs. B. Stabilize below program level due to only having one MBFP in service. The AOP provides guidance to address this by controlling the MBFP and FRVs in manual as necessary to return level to program. C. Increase to above program and then return to program level following a damped oscillation" The AOP does not provide guidance for any actions since level returns to program. D. Increase and stabilize above program due to integral error signal induced by the transient. The AOP provides guidance to address this by controlling the MBFP and FRVs in manual as necessary to return level to program.

Answer: A Explanation/Justification: This question requires the candidate to know that the controllers for the FRVs are going to "windup" due to level being below program for an extended period of time. AOP-FW-1 would have the operators remove the "windup" from the FRVs, The candidate must remember this action and understand why it is done. With no actions, a Turbine trip will occur due to high level on SGs from overfeeding. Eliminating distractors will be easier if the candidate has a strong understanding of how the control system responds to being off program. A. Increase until a Turbine Trip occurs due to integral error in the feedwater control system. To prevent this, the AOP will remove the integral error from the FRVs. B. Stabilize below program level due to only having one MBFP in service. The AOP will address this by controlling the MBFP and FRVs in manual as necessary to return level to program. C. Increase to above program and then return to program level following a damped oscillation. The AOP does not require any actions since level returns to program. D. Increase and stabilize above program due to integral error signal induced by the transient. The AOP will address this by controlling the MBFP and FRVs in manual as necessary to return level to program. Technical

References:

2-AOP-FW-1 Proposed References to be provided: None Learning Objective: 12LP-ILO-ICSGL - 5 Question Source: Bank # ---- IPEC Bank Note changes or Modified Bank #

                                                 - - - - attach parent New                      x Question History:                    Last 2 NRC Exams at IPEC:               NA Memory or Fundamental Question Cognitive Level:            Knowledge:

Comprehension or Analysis: x

Exam Outline Cross

Reference:

Level SRO Tier# 90 Group # 1 KJA# 103000A203 Ability to (a) predict the impacts of the following malfunctions or operations on the Containment System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation Importance 3.8 Question # 90 The following plant conditions exist on Unit 2: Reactor is at 100% RTP

     -  A manual Phase A isolation signal was inadvertently actuated on Train A.

Which of the following are direct results of this signal and corrective actions required to be taken in response to this event? Results Actions LCV-459, Letdown Isolation Loop Reset Phase A A. 21, and all orifice isolation valves Restore Instrument Air to VC close Place Excess Letdown in service LCV-459, Letdown Isolation Loop Reset Phase A B. 21 and all orifice isolation valves Restore Instrument Air to VC close Place Letdown in service 201, "Isolation Valve Letdown Line Reset Phase A Normal Path" C. Restore Instrument Air Isolation," and all orifice isolation valves close. Place Letdown in service. 201, "Isolation Valve Letdown Line Normal Path" Reset Phase A D. Isolation" and all orifice isolation Restore Instrument Air to VC I I valves close. Place Excess Letdown in service

Answer: C Explanation!Justification: A. Incorrect. Plausible because 459 does not isolate on a Phase A signal. There is no reason to place excess letdown in service if normal letdown is available. B. Incorrect; Plausible because 459 does not isolate on a Phase A signal. The actions for this distractor are correct. C. Correct. 201 and all orifice isolation valves do isolate on a Phase A signal. Since normal letdown is available, this would be preferred to excess letdown. D. Incorrect. Plausible because the Results are correct but the actions are not. There is no reason to place excess letdown in service if normal letdown is available. 2-AOP-CVCS-1 Technical

References:

2-PT-R141 Proposed References to be provided: None Learning Objective: 12LP-ILO-CVCS - 5 Question Source: Bank# IPEC Bank Note changes or Modified Bank # attach parent New x Question History: Last NRC Exam: Memory or Fundamental

                                                         -NA- - -

Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _---'(--I.b)'--_~ _ __ 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level RO Tier # Group # 2 KlA# 045000A217-------:~~- Ability to (a) predict the impacts of the following malfunctions or operations on the MT/G System and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Malfunction of electrohydraulic control Importance 2.9 Question # 91 The plant is at 50% power. A malfunction has occurred resulting in an 8 psig increase in governor oil pressure. Which of the following correctly identifies the impact of this malfunction and the required operator actions as specified in 2 AOP-LOAD-1, Excessive Load Increase or Decrease? Answer: B Explanation/Justification:

Governor Oil pressure increases from approximately 20 psig to approximately 40 psig from Latched to full power. At 50% power the pressure is approximately 28 psig. When power is < 75% the load limits are maintained within 5 psig of governor oil pressure. When the governor oil pressure increases above Load Limit oil pressure, the Load Limit oil pressure will be in control. The AOP directs turbine load reduction to restore Tavg not rod adjustment. A. Incorrect. Plausible because the aux governor will limit load if it is increasing at 3% per second. If 2 PR Nls exceed 108% the reactor should be tripped; however a 5 psig (Maximum increase) from 50% will not cause power to exceed 108% B. Correct C. Incorrect. Plausible because the load will be limited by the load limit setpoint; however, restoring Tavg using control rods is not desired and may make ~I worse without a boron adjustment. D. Incorrect. Plausible because the candidate must remember that the load limit oil pressures are maintained within 5 psig of control oil pressure. Furthermore the candidate must remember that the lower of the oil pressures (governor or load limit) controls the turbine. If 2 PR Nls exceed 108% the reactor should be tripped; however an 5 psig (Maximum increase) from 50% will not cause power to exceed 108%. Technical

References:

2-AOP-LOAD-1

                                            ~~~~~~--------

Proposed References to be provided: _N_on_e Learning Objective: 12LP-ILO-MTG001 - 7 12LP-ILO-MTG001 - 5 Question Source: Bank # - - - - IPEC Bank Note changes or Modified Bank # - - - - attach parent New x Question History: Last 2 NRC Exams at IPEC: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: x 10 CFR Part 55 Content: 55.41 55.43 5 Comments:

Exam Outline Cross

Reference:

Level SRO Tier # 2 Group # 2 KIA # 0620002240 Equipment Control - Ability to apply technical specifications for a system. Importance 4.7 Question # 92 The normal supply breaker to 480V Bus 3A opens due to a breaker malfunction. The crew responds per 2-AOP-480V-1, Loss of Normal Power to Any 480V Bus, and is unable to close breaker EG-2B. There is no damage to 3A and all fault indications are cleared. Based on these conditions, what actions will be directed by 2-AOP-480V-1? A. Regardless of mode, no actions will be taken to power any of the loads normally fed from 480V Bus 3A until the bus can be powered from either its normal feed or 22 EDG. B. Rack in and close breaker 2AT3A if RCS temperature is < 200°F since there is no damage to Bus 3A. C. Rack in and close breaker 2AT3A if RCS temperature is < 350°F since there is no damage to Bus 3A. D. Regardless of mode, rack in and close breaker 2AT3A since there is no damage to Bus 3A and declare 480V Safeguards Busses 2A13A inoperable. Answer: B Explanation/Justification: /' <... A. Incorrect but plausible. The procedure will have~r backup supplied to 23 Battery Charger. Also 2AT3A can be closed if temperature is <200F. It is plausible because the AOP may not ever specifify tying the breakers. B. Correct. Based on 2-AOP-480V-1 Attachment 2 step 2.160/2.161

C. D. Incorrect but plausible. This is plausible because closing the tie breaker would only jeopardize the 2A and 3A busses. Since the answer states that these busselS"would be declared inoperable, it is plausible that the procedure could specify this. 2-AOP-480V-1 Att 2 Technical

References:

Tech Specs 3.8.2 Basis Proposed References to be provided: None Learning Objective: 12LP-ILO-AOP480 - 2 Question Source: Bank# ---- IPEC Bank Note changes or Modified Bank # attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _--->(--'b)'--7_ _ _ __ 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level Tier # Group # 2 KIA #

                                                   ...... ILV to   predict the impacts of the following malfunctions or operations on the RPIS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Loss of power to the RPIS Importance 3.6 Question # 93 The unit is operating at 100% power. The following annunciators are in alarm:

Approaching Rod Insertion Limit 1.25"
Rod Insertion Limit 0"
Rod Control Non Urgent Failure
Rod Bottom Rod Stop AlllRPls indicate 0" and all rod bottom lights are extinguished.

What event has occurred and what actions are required? EVENT ACTIONS I i A. MCC-24 is de-energized Be in MODE 3 in 6 hours Place control rods under manual B. MCC-24 is de-energized control 23 Instrument Bus is de- Reduce THERMAL POWER to S 75% C. energized RTP 23 Instrument Bus is de- Verify SDM to be within the limits D. energized specified in the COLR Answer: B Explanation!Justification:

A. Incorrect. Plausible because MCC-24 is the power supply to the IRPI. TS action is not correct. B. Correct. MCC-24 is the power supply to the IRPI. All ala~s and indications are consistent with a loss of power. Placing rord control in manual satisfies TS 3.1.7. C. Incorrect. Plausible because 23 Instrument bus supplies most of the indications and controllers on the flight panel. Reducing thermal power to

     <75% is a TS action for a misaligned rod.

D. Incorrect. Plausible because 23 Instrument bus supplies most of the indications and controllers on the flight panel. Verification of SDM is TS action for misaligned/dropped rods. Tech Spec 3.1.7 Technical

References:

2-AOP-480V-1 Proposed References to be provided: None Learning Objective: 12LP-ILO-ICRPI - 7 12LP-ILO-ICRPI-14 Question Source: Bank# IPEC Bank

                                               ----         Note changes or attach Modified Bank #                  parent New                     x Question History:                   Last NRC Exam:         NA
                                                        ---'--'-'-,'--~-

Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 ,_-.l(..b)L-_ 55.43 (b) 2 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Conduct Group # _ofOps KIA # 1940012107 Conduct of Operations - Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. Importance 4.7 Question # 94 While determining if RCPs should be stopped in ECA-2.1, Uncontrolled Depressurization of All Steam Generators, the following plant conditions exist:

Reactor Coolant System Pressure is 1035 psig and stable.
Hot Leg temperature - 340°F and stable.
Cold Leg temperature - 325°F and stable.
Subcooling - 200°F.
All Reactor Coolant Pumps running.
Steam Generator Pressures 2.L 22 23 24 250 psig 230 psig 230 psig 250 psig increasing decreasing decreasing stable Which ONE of the following is the correct course of action?

A. Transition to FR-P.1, Response to Imminent Pressurized Thermal Shock Condition B. Continue in ECA-2.1, Uncontrolled Depressurization of All Steam Generators C. Transition to E-2, Faulted Steam Generator Isolation D. Transition to E-3, Steam Generator Tube Rupture Answer: C

Explanation!Justification: A. Incorrect but plausible since FR-P.1 criteria are B. Incorrect but plausible since there are situati where transition to E2 is delayed when a MSIV is closed. ((Y ~ ct but plausib~here are si d ayed when a<M'SlV is closed. ns wh~n to E2 is D. I correct but plausible since an operator could mistake the increasing SG pressure as being due to a SGTR. Technical

References:

2-ECA-2:t Foldout Page Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPC21 - 1 Question Source: Bank#

                                               - -x- -       IPEC Bank 15864 Note changes or Modified Bank #     _ _ _ _ attach parent New Question History:                   Last NRC Exam:        NA
                                                          -~--

Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 5 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Conduct Group # ofOps KIA # 1940012145 Conduct of Operations - Ability to identify and interpret diverse indications to validate the response of another indication. Importance 4.3 Question # 95 Given:

Unit 2 has experienced a large break LOCA.
All safeguards equipment operated as designed.
The crew has transitioned to 2-ES-1.3.
One RWST Low Low Level alarm is illuminated.

Which of the following is used to determine if a level transmitter has failed? A. Compare RWST level to Containment level B. Check time from SI initiation> 30 minutes C. Check Containment Sump level increasing D. Check Containment Sump level> 46' 8 Yz " Answer: 0 Explanation/Justification: A. Incorrect. Plausible because ECA-3.1 (Not ES-1.3) has a graph to compare RWST level with expected Containment level. B. Incorrect. Plausible because the length of time to reach the low low level setpoint is approximately 25 minutes C. Incorrect. Plausible because checking the level increasing is done if both RWST low low level alarms are illuminated; however to confirm adequate level for recirc/RHR pump operation requires checking containment sump level (Le., actual sump level may be inadequate for pump NPSH).

D. Correct. When RWST is at approximately 9.24', containment level should be approximately 46' 9 % " to provide adequate NPSH for the recirc/RHR pumps. Technical

References:

2-ES-1.3 Background Proposed References to be provided: None Learning Objective: 12LP-ILO-EOPS13 - 4 Question Source: Bank# IPEC Bank Note changes or Modified Bank # ---- attach parent New x Question History: Question Cognitive Level: Last NRC Exam: Memory or Fundamental Knowledge: NA x Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Equip Group # Control KIA # 1940012225 Equipment Control Knowledge of bases in technical specifications for limiting conditions for operations and safety limits. Importance ~ 4.2 Question # 96 ~ 1) A malfunction of the Master Pressurizer Level Co roller occurred which resulted in PZR Level reaching the Technical Specificatio iJi..rnit. The basis for this limit is to... I'l'-... A. ensure that the PZR PORVs will be able to respond to a PZR insurge and prevent a High Pressure Reactor Trip from occurring. B. ensure that the High Pressure Reactor Trip can prevent lifting of PZR Code Safety Valve. C. ensure that Normal PZR Spray will remain capable of preventing a High Pressure Reactor Trip following a 50% load rejection with HP Steam Dumps available. D. ensure safety analyses assumptions are met for events that result in PZR insurge (e.g. loss of normal feedwater). Answer: D ExplanationlJustification: A. Incorrect but plausible. Part of the statement is correct (the PORVs do open before a high presure trip). but this is not a true statement because PZR level does not come into play for this. B. Incorrect but plausible. This is not the basis for PZR level. This is plausible because the High Pressure Trip does prevent lifting the safeties. C. Incorrect but plausible. This would be a true statement for purpose of spec on PORVs D. Correct. See T.S. Bases 3.4.9 applicable safety analyses section. Technical

References:

Tech Spec 3.4.9 Basis

Proposed References to be provided: Learning Objective: 12LP-ILO-RCSPZR - 14 Question Source: Bank# IPEC Bank Note changes or Modified Bank # attach parent New x Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: X Comprehension or Analysis: 10 CFR Part 55 Content: 55.41 _ _~{--,b),--5_ _ _ __ 55.43 (b) 2 Comments:

Exam Outline Cross

Reference:

Level Tier# Generic Equip Group # Control KIA # 1940012221 Equipment Control Knowledge of pre- and post maintenance operability reqUirements. Importance 4.1 Question # 97 Given the following conditions:

Maintenance requested a tagout for 2381 Pump
The tagout included placing the control room switch in pullout and racking out the supply breaker only
A visual inspection showed no work was required
No disassembly work was performed Which ONE of the following indicates the minimum requirement for restoring 23 81 Pump operability?

23 81 Pump can be considered operable when: A. Its supply breaker is racked in and its control room switch is back in AUTO B. Its supply breaker is racked in and its control room switch is back in AUTO and the pump has been started using the control room switch. C. Its supply breaker is racked in and its control room switch is back in AUTO and the pump has been started using the control room switch and verified to meet the acceptance criteria of the quarterly surveillance test. D. Its supply breaker is racked in and its control room switch is back in AUTO and the pump has been started using the control room switch AND using the auto-start relay. Answer: 8 Explanation/Justification: A. Incorrect but plausible. A candidate may believe that once the breaker is racked in that the pump would auto start.

B. Correct. The minimum requirement is to have the equipment in the proper configuration and verify the breaker will close and start the pump C. Incorrect but plausible. It is not necessary to verify the pump develops head etc. because nothing that was done calls in to question that this was affected. This is plausible because we generally schedule the routine surveillance to occur after removing a PTO for maintenance D. Incorrect but plausible. It is not necessary to verify that the pump auto start circuitry works. A candidate could conclude it is because racking out the breaker may affect breaker cell switches requiring testing auto-start. Technical

References:

OAP-37 Proposed References to be provided: -'N--'-o-'n_e=--________________________ Learning Objective: IOLP-ILO-ADM01 Question Source: Bank# - - - - IPEC Bank Note cha nges or Modified Bank #

                                                ---- x        attach parent New Kewaunee Question History:                      Last NRC Exam:       -2002 Memory or Fundamental Question Cognitive Level:              Knowledge:

Comprehension or Analysis: x 10 CFR Part 55 Content: 55.41 (b) 55.43 (b) 2 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Rad Group # Controls KIA # 1940012314 Radiological Controls Knowledge of radiation or contaminatin hazards that may arise during normal, abnormal, or emergency conditions or activities. Importance 3.8 Question # 98 Which of the following activities is likely to lead to highest levels of surface contamination if equal amounts of liquid were to leak on the floor? A. Disconnecting Reactor Cavity drain hoses following refueling operations. B. Removing a test gauge from Safety Injection Pump discharge piping while online. C. A leaking fitting on outlet piping of flush water for a depleted Cation bed. D. A leaking CVCS Seal Injection Filter housing while online. Answer: A Expla nation!Justification: A. Correct. Generally, the highest levels of contamiation would be expected from the Reactor Cavity following refueling. B. Incorrect but plausible. Contamination levels should be low, but a candidate may wonder if RCS water would come out of discharge piping. The dose rate from online RCS may be higher than cavity drain water, but particulates are much less. C. Incorrect but plausible. Actual contamination levels from spent resin activities are generally low because the resin holds a lot of the particulates. A candidate may believe that because a lot RCS volume passed through the bed that contamination levels would be high.

D. Incorrect but plausible. Contamination levels should not be that bad in CVCS (has gone through demins and reactor filter) compared to Cavity drain water. Because this water would online RCS a candidate may think levels would be high. Also the candidate is not told if the water is up or downstream of the filter. Technical

References:

Proposed References to be provided: None Learning Objective: Question Source: Bank# IPEC Bank

                                            -----         Note changes or Modified Bank #    _______ attach parent New                     x Question History:                  Last NRC Exam:
                                                        -NA Memory or Fundamental Question Cognitive Level:          Knowledge:

Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _---->.(b----t)_ _ _ __ 55.43 (b) 4 Comments:

Exam Outline Cross

Reference:

Level Tier # Generic Emerg Group # Proc/Plan KIA # 940012404 Emergency Procedures/Plan Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emergency and abnormal operating procedures. Importance 4.7 Question # 99 The following plant conditions exist:

The plant was operating at 100% power
The Team has initiated a manual reactor trip and Safety Injection
RCS pressure is 1650 psig and slowly decreasing
SG pressures are all stable at 900 psig
No SG Level is increasing in an uncontrolled manner
Auxiliary feedwater flow is 450 gpm and stable
All SI pumps are running
Containment Radiation levels, CNMT Temperature and pressure remain normal
Primary Auxiliary Building radiation levels are increasing
All secondary side radiation monitor readings are normal If the RCS leakage cannot be isolated, which ONE of the following EOP procedure sequences would be utilized to address these conditions upon transition from E-O Reactor Trip or Safety Injection?

E-1, Loss of Reactor or Secondary Coolant ECA-1.2, LOCA Outside of Containment ECA-1.1, loss of Emergency Coolant Recirculation A. E-1 to ECA-1.2 to E-1. B. ECA-1.2 to ECA-1.1. C. E-1 to ECA-1.2 to ECA-1.1. D. ECA-1.2 to E-1.

Answer: B Explanation!Justification: A. Incorrect but plausible because an operator may not know that E-1 is not entered prior to ECA-1.2 and that E-1 is not where the team goes upon exit from ECA-1.2 if break is not isolated. B. Correct C. Incorrect but plausible because may not know that E-1 is not entered prior to ECA-1.2. D. Incorrect but plausible because an operator may not know that E-1 is not where the team goes upon exit from ECA-1.2. Technical

References:

                    -==~~2::=:-E=-C~A~-1~.2~_ _ _ _ _ __

Proposed References to be provided: _N~o.=..:n-,-,e,,--_ _ _ _ _ _ _ _ _ __ Learning Objective: 12LP-ILO-EOPC12 - 5 Question Source: Bank# ---- x IPEC Bank 8315 Note changes or Modified Bank # ---- attach parent New Question History: Last NRC Exam: NA Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 (b) 10 55.43 (b) 5 Comments:

Exam Outline Cross

Reference:

Level Tier# Generic Emerg Group # Proc/Plan KIA # 1940012430 Emergency Procedures/Plan Knowledge of which events related to system operations/status that must be reported to internal orginazations or external agencies, such as State, the NRC. or the transmission system operator. Importance 4.1 Question # 100 The plant is in an outage in Mode 5 when a complete loss of RHR occurs. Temperature increases and is stabilized at 220°F by the SG Atmo$pheric Dump Valves throttling. Which ONE of the following identifies when the NRC is required to be notified of this event? Notify the NRC within ... A. 1 hour B. 4 hours C. 8 hours D. 30 days Answer: A Explanation/Justification: EAL 8.2.3 met. E-Plan declaration requires 1 hour report to NRC A. Correct B. Incorrect but plausible. The other available report times are all valid for various equipment failures.

C. Incorrect but plausible. The other available report times are all valid for various equipment failures. D. Incorrect but plausible. The other available report times are all valid for various equipment failures. Technical

References:

IP-SMM-U-108 Proposed References to be provided: None Learning Objective: IOLP-ILO-ADM01 - 1 Question Source: Bank# ----- IPEC Bank Note changes or Modified Bank # ----- x attach parent New Nine Mile Question History: Last NRC Exam: Point 2002 Memory or Fundamental Question Cognitive Level: Knowledge: Comprehension or Analysis: X 10 CFR Part 55 Content: 55.41 _ _------>.(b.::....L)_ _ _ __ 55.43 (b) 5 Comments:}}

v t e Letter Sequence Draft Other
TAC:U01787, (Open)
Initiation Acceptance... Administration Questions Answers Results Approval... Other: ML102510238, ML102640428, ML102640438, ML102640564, ML102640579, ML102640589, ML102640597
MONTHYEAR2010-06-22 [Table View]

ML102160334: Difference between revisions

Latest revision as of 20:31, 11 March 2020

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