ML15076A390

2015 Point Beach Nuclear Plant Initial License Examination Proposed Written Examination

ML15076A390
Person / Time
Site:	Point Beach
Issue date:	03/06/2015
From:	Mcneil D NRC/RGN-III/DRS/OLB
To:
Shared Package
ML14328A661	List: ML15070A500 ML15070A522 ML15070A537 ML15070A540 ML15076A390 ML15076A507 ML15076A510 ML15076A514 ML15076A517 ML15076A521 ML15076A525 NRC-2014-0087, 2015 Point Beach Nuclear Plant Initial License Examination Proposed Examination Transmittal Letter
References
50-266/15-301, 50-301/15-301
Download: ML15076A390 (199)

Text

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 1 of 199 QUESTION # 001 Given the following:

• Unit 1 is operating at Rated Thermal Power

• A leak equivalent to 1/2 inch in diameter has developed on the line connecting the pressurizer to the PORVs Which of the following describes the INITIAL response of the pressurizer pressure control and level control systems? (Assume no operator action)

Pressurizer Pressure Controller Charging Pump in Auto 1HC-431K 1HC-428A A. output will RISE (towards 100%) will speed up B. output will RISE (towards 100%) will slow down C. output will LOWER (towards 0%) will speed up D. output will LOWER (towards 0%) will slow down ANSWER:

D.

REFERENCE:

883D195 Sh 18, Logic Diagram Pressurizer Pressure and Level Control, Rev 13 MODIFIED: 2012 Sequoyah NRC Exam HIGHER

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 2 of 199 K/A: 008AK2.03 Pressurizer (PZR) Vapor Space Accident (Relief Valve Stuck Open) Knowledge of the interrelations between the Pressurizer Vapor Space Accident and the following: Controllers and positioners.

LEARNING OBJECTIVE:

DESCRIBE the automatic functions and interlocks associated with the Pressurizer Level Control, Pressure control, and Relief System and its major components: a. Pressurizer Pressure b. Pressurizer Level (051.01.LP0078.004)

EXPLANATION:

A. INCORRECT: See (d.)

B. INCORRECT: See (d.)

C. INCORRECT: See (d.)

D. CORRECT: Pressurizer pressure will be dropping due to the vapor space LOCA. This will cause system pressure to lower, as pressure drops below the setpoint, the output of the controller will lower to turn on the heater and raise pressure back to the setpoint. The drop in pressure will allow the pressurizer level to rise as additional water enters the pressurizer via the surge line. This level rise will cause the charging pump in auto to slow.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 3 of 199 Original Question:

Given the following plant conditions:

• Unit 1 is operating at 100% power.

• A leak equivalent to 1/2 inch in diameter has developed on the line connecting the pressurizer to the PORVs.

Which ONE of the following describes the long term response of the pressurizer pressure control and level control systems?

Pressurizer Pressure Controller Pressurizer Level Controller 1 -PIC-68-340A 1 -LIC-68-339 A. output will INCREASE output will INCREASE B. output will INCREASE output will DECREASE C. output will DECREASE output will INCREASE D. output will DECREASE output will DECREASE

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 4 of 199 QUESTION # 002 Given the following:

• Unit 1 has experienced a plant transient that resulted in an automatic reactor trip and safety injection from 50% reactor power.

• The crew is responding with EOP-0, Reactor Trip or Safety Injection.

• The reactor operator notes that both Main Steam Isolation Valves (MSIVs) have automatically shut.

Assuming all equipment functioned as designed, which of the following initiating events could have caused BOTH MSIVs to automatically shut and why?

A. (1) Steam line A faulted due to a failed-open safety valve.

(2) The MSIV closure prevents excessive cooldown due to both steam generators blowing down.

B. (1) Steam line B faulted immediately downstream of 1MS-2018B, MSIV, in the façade.

(2) The MSIV closure prevents excessive cooldown due to either steam generator blowing down.

C. (1) Steam generator tube rupture on both steam generators.

(2) The MSIV closure prevents uncontrolled release of radioactivity from the containment.

D. (1) Large break LOCA which results in Containment Spray actuation.

(2) The MSIV closure prevents potential release of radioactivity from containment.

ANSWER:

D.

REFERENCE:

FSAR Chapter 5.2 Containment Isolation Signals, Rev 2014 FSAR Chapter 7, Steam Line Isolation Signals, Rev 2014 FSAR Table 7.3.1 ESFAS Steam Line Isolation Signals, Rev 2014 STPT 2.2, Setpoint Document, Steam Line Isolationm, Rev 6 NEW HIGHER K/A: 011EK3.01: Knowledge of the reasons for the following responses as they apply to the Large Break LOCA: Verifying main steam isolation valve position.

Learning Objective:

EXPLAIN the ECCS response to a large break LOCA.

(043.03.LP2463.008)

EXPLANATION:

A. INCORRECT: If the A main steam line faulted at the location of the safeties, the non-return check valve in the A main steam line would prevent the B generator from blowing down through the leak. Only the

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 5 of 199 A MSIV would have likely closed on high steam flow coincidence.

Second part is plausible in that both MSIVs are shut in standard WOG procedure set as no check valve is in the standard plant.

B. INCORRECT: If the B main steam line faulted in the façade downstream of the MSIV, the non-return check valve in the B main steam line would prevent the A generator from blowing down through the leak. Only the B MSIV would have likely closed on high steam flow coincidence. Second part is true in that if both MSIVs were to close with a break in this location, the neither SG would fully blow down.

C. INCORRECT: While closure of both MSIVs would prevent uncontrolled release of radioactivity, the steam generator tube rupture transient would not result in automatic MSIV closure.

D. CORRECT: The high-high containment pressure setpoint of 15 psig has been exceeded, causing both MSIVs to automatically close. MSIV closure prevents radioactivity release from containment should the main steam system also be damaged inside containment.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 6 of 199 QUESTION # 003 Given the following:

At 10:00, Unit 1 plant conditions were as follows:

• The Unit is operating at Rated Thermal Power

• The following parameters are observed:

o Letdown flow: 40 gpm o Charging line flow: 25 gpm o Seal Injection flow: 9.4 gpm per RCP o Seal return flow: 2.2 gpm per RCP At 12:00, Unit 1 plant conditions were as follows:

• The Unit is operating at Rated Thermal Power

• 1RC-427, RC Loop B Cold Leg Letdown Isolation, has failed closed

• Normal letdown cannot be established

• Excess letdown has been placed in service

• The following parameters are observed:

o Excess Letdown flow has been determined to be 12 gpm o Charging line flow: 0 gpm o Seal injection flow: 9.4 gpm per RCP o Seal return flow: 2.2 gpm per RCP Which of the following identifies the trend in Pressurizer level at 10:00 and 12:00?

Pressurizer level at 10:00 was and Pressurizer level at 12:00 was A. Increasing Increasing B. Increasing Decreasing C. Decreasing Increasing D. Decreasing Decreasing ANSWER:

C.

REFERENCE:

LP0079, Chemical and Volume Control, Rev 24 MODIFIED: 2012 Sequoyah NRC Exam HIGHER K/A: 022AA1.03: Ability to operate and / or monitor the following as they apply to the Loss of Reactor Coolant Makeup: PRZ level trend.

LEARNING OBJECTIVE:

PREDICT and ASSESS the effect(s) of the following operations/events:

a. Loss of CVCS on other systems

b. Changing CVCS orifice configuration EXPLANATION:

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 7 of 199 A. INCORRECT: See distractor (c.)

B. INCORRECT: See distractor (c.)

C. CORRECT: Total letdown flow = letdown flow + RCP seal return flow.

Total charging flow = charging line flow + RCP seal injection flow. The difference in total letdown flow when compared to total charging flow will determine whether or not the pressurizer is trending up or down in level. In the first half of the question, the total letdown flow is 44.4 gpm and the total charging flow is 43.8 gpm. Letdown is greater than charging, and therefore pressurizer level is trending down. In the second half of the question, total letdown flow is 16.4 gpm and the total charging flow is 18.8 gpm. Charging is greater than letdown, and therefore pressurizer level is trending up.

D. INCORRECT: See distractor (c.).

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 8 of 199 Original Question:

Given the following plant conditions:

AT 10:00, Unit 1 plant conditions were as follows:

• 100% steady state power.

• The following parameters are observed:

o Letdown flow: 75 gpm o Total charging flow: 100 gpm o Seal Injection flow: 10 gpm per RCP o Seal return flow: 3 gpm per RCP At 12:00, Unit 1 plant conditions were as follows:

• 100% steady state power.

• 1-FCV-62-93, Charging Flow Control Valve, has failed closed.

• Normal letdown cannot be established.

• Excess letdown has been placed in service.

• The following parameters are observed:

o Excess Letdown flow: at maximum capacity of heat exchanger:

25 gpm o Total charging flow: 40 gpm o Seal injection flow: 10 gpm per RCP o Seal return flow: 3 gpm per RCP Which ONE of the following identifies the trend in Pressurizer level at 10:00 and 12:00? Pressurizer level at 10:00 was _(1)_ and pressurizer level at 12:00 was _(2)_.

(1) (2)

A. Increasing Increasing B. Increasing Decreasing C. Decreasing Increasing D. Decreasing Decreasing

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 9 of 199 QUESTION # 004 Given the following:

• Unit 1 is operating at Rated Thermal Power Concerning the Component Cooling Water (CCW) system, a (1) will procedurally require the crew to (2) to prevent damage to components.

A. (1) Thermal Barrier leak (2) immediately trip and stabilize the plant using EOP-0, Reactor Trip or Safety Injection, then stop the RCPs B. (1) RCP Motor Bearings cooler leak (2) verify seal injection supply, isolate cooling flow to the oil coolers, monitor RCP operation C. (1) Seal Return Heat Exchanger leak (2) commence a rapid load reduction, due to a loss of cooling to the VCT D. (1) Surge Tank level at 5%

(2) immediately place CCW pumps in pullout, then trip and stabilize the plant using EOP-0, Reactor Trip or Safety Injection, then stop the RCPs ANSWER:

D.

REFERENCE:

AOP-9B, Component Cooling System Malfunction, Step 3, Rev 22 BG AOP-9B, Background Document Component Cooling System Malfunction. Rev 16 NEW HIGHER K/A: 026AA2.06: Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: The length of time after the loss of CCW flow to a component before that component may be damaged.

LEARNING OBJECTIVE:

Given access to the Site Specific Simulator or specific plant conditions, DEMONSTRATE the ability to maintain CCW Surge Tank level and pressure. (055.03.LP2444.003)

EXPLANATION:

A. INCORRECT: The correct response for a thermal barrier leak is to ensure seal injection supply, and isolate CCW to the thermal barrier.

Plausible because, the student must recognize this as a CCW malfunction not an RCS leak, and remember that interlocks are used to mitigate this event. This course of actions are also called out in the AOP for Loss of CCW.

B. INCORRECT: The correct response for a motor bearing cooler leak is to trip and stabilize the plant, and then trip the RCPs. Plausible

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 10 of 199 because these are the required actions for a thermal barrier leak.

Plausible because the student needs to differentiate between the two CCW/RCP malfunctions.

C. INCORRECT: The correct response for this is to bypass the seal return heat exchanger and if a VCT high temperature alarm comes in, trip the plant. Plausible, because no value for the leak was given, therefore a rapid load reduction would seem logical.

D. CORRECT: These are the correct actions to take on a surge tank level less than 10%. It is to protect the pumps from a loss of flow condition caused by low NPSH.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 11 of 199 QUESTION # 005 Given the following:

• An ATWS has occurred on Unit 2

• The crew is performing the actions in accordance with CSP-S.1, Response to Nuclear Power Generation/ATWS

• Boration flow was initiated by starting a boric acid transfer pump and positioning 2CV-350, Emergency Boration BA to CHG Pump MOV, to OPEN

• PZR pressure is 2390 psig

• Charging Line Flow is 28 gpm

• VCT level is Rising What action is the operator required to take next?

A. Start an additional Boric Acid Pump B. Start a third Charging Pump and establish maximum charging C. Close 2CV-112C, the VCT Outlet to CHG Pump Suction valve D. Ensure PZR PORV(s) and block valves are open, until PZR pressure is less than 2335 psig ANSWER:

D.

REFERENCE:

CSP-S.1, Response to Nuclear Power Generation/ATWS, Rev 37 BG CSP-S.1, Background Document Response to Nuclear Power Generation/ATWS, Rev 26 BANK: 2009 Seabrook NRC Exam HIGHER K/A: 029G2.1.20: Anticipated Transient Without Scram (ATWS)

Ability to interpret and execute procedure steps.

LEARNING OBJECTIVE IMPLEMENT the CSPs to respond to plant conditions where the Subcriticality Status Tree is not satisfied. (043.03.LP1996.013)

EXPLANATION:

A. INCORRECT: The Boric Acid Transfer pump indicates that it is running based on the increase in VCT level. A second Boric Acid Transfer Pump is not needed by procedure, and if the pumps were incapable to delivering flow (one could not be started) then the procedure will have the operator line up charging from the RWST.

Plausible because the step allows the use of either pump and the student must determine the pump is running by the increase in VCT level.

B. INCORRECT: Starting a second charging pump and charging at maximum would only be required if borating the primary from the RWST, and that has not been performed. Plausible because this will be done if boration is being supplied from the RWST.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 12 of 199 C. INCORRECT: Closing this valve is half of the required action to supply boration from the RWST, and should not be done at this time.

Plausible if the student believes this will assist in the delivery of boron to the primary.

D. CORRECT: This is the correct action based on the primary pressure.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 13 of 199 QUESTION # 006 Given the following:

• A Steam Generator Tube Rupture has occurred

• The crew is currently in EOP-3, Steam Generator Tube Rupture

• The crew has just completed the RCS cooldown to the target temperature of 495°F based on Core Exit Thermocouples The OS has just asked you to depressurize the RCS while maintaining a minimum of 35°F subcooling.

At the current RCS temperature, which of the following is the lowest that RCS pressure can be without violating the subcooling requirement?

A. 900 psig B. 885 psig C. 871 psig D. 651 psig ANSWER:

C.

REFERENCE:

EOP-3 Unit 1, Steam Generator Tube Rupture, Step 17, Rev 45 Steam Tables Proposed reference to be provided to the applicants during examination:

Steam Tables MODIFIED: 2008 Beaver Valley NRC Exam HIGHER K/A: 038EK1.01: Knowledge of the operational implications of the following concepts as they apply to the SGTR: Use of steam tables.

LEARNING OBJECTIVE:

Given access to Site Specific Simulator, IMPLEMENT the EOPs to respond to a faulted ruptured Steam Generator.

(031.02.LP0441.016)

EXPLANATION:

A. INCORRECT: 495°F + 35°F = 530°F calculates to 884.82 psia mistakenly adding 14.7 = 899.52 psig.

B. INCORRECT: 495°F + 35°F = 530°F calculates to 884.82 psia C. CORRECT: 495°F + 35°F = 530°F calculates to 884.82 psia - 14.7 =

870.12 psig.

D. INCORRECT: 495°F saturation temperature is 650.41 psia

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 14 of 199 Original Question:

• The plant is operating at 100% power with all systems .... in NSA.

• A Steam Generator Tube Rupture occurs.

• The crew enters the EOP network.

• The crew is currently implementing E-3, Steam Generator Tube Rupture.

• The RCS has been cooled to 500°F in preparation for equalizing RCS pressure with the ruptured SG pressure.

• The Unit Supervisor directs you to depressurize the RCS AND while maintaining a minimum of 20°F of Subcooling.

At the current RCS temperature, what is the lowest RCS temperature can be without violating the 20°F of Subcooling requirement?

A. ~ 666°F B. ~ 695°F C. ~ 798°F D. ~ 827°F

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 15 of 199 QUESTION # 007 Given the following:

• Unit 1 was at Rated Thermal Power

• Twenty (20) minutes ago, an RCP flywheel catastrophically failed

• AFW flow of 115 gpm per Steam Generator was established in EOP-0, Reactor Trip or Safety Injection

• ECA-2.1, Uncontrolled Depressurization of Both Steam Generators, feed control step is in progress

• SG 'A' Wide Range level is 200 inches and lowering

• SG 'B' Wide Range level is 100 inches and lowering

• RCS temperature is now 400°F and lowering Which of the following actions will the crew take regarding AFW flow to the Steam Generators per ECA-2.1 and what is the reason for this action?

A. Maintain AFW flow at 115 GPM to each SG, since 230 GPM flow is required to maintain adequate heat sink.

B. Secure AFW flow to 'B' SG and raise AFW flow to 'A' SG to 230 GPM to isolate feed to the more severely faulted SG and maintain adequate heat sink.

C. Secure AFW flow to 'B' SG and maintain AFW flow to 'A' SG at 115 GPM since 115 GPM flow is directed to minimize RCS cooldown and 'B' SG level is lowering faster.

D. Reduce AFW flow to 50 GPM to each SG to minimize cooldown while maintaining both SGs in a wet condition.

ANSWER:

D.

REFERENCE:

ECA-2.1, Uncontrolled Depressurization of Both Steam Generators, Step 2, Rev 41 BANK: 2012 PBNP NRC Exam (Last 2 NRC Exams)

HIGHER K/A: W/E012EK1.01 - Knowledge of the operational implications of the following concepts as they apply to the (Uncontrolled Depressurization of All Steam Generators): Components, capacity, and function of emergency systems.

LEARNING OBJECTIVE STATE the entry conditions and IDENTIFY the major actions for the following procedures:

d. ECA 2.1. (031.02.LP0465.001)

EXPLANATION:

A. INCORRECT: ECA-2.1 directs lowering feed flow if cooldown of 100F/Hr has been exceeded. Although RED path for heat sink will be indicated when flow is lowered, the operator will be in control of

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 16 of 199 temperature and CSP-H.1 will not need to be entered. If cooldown rate has not exceeded 100F/hr, if cooldown was not excessive, this would be correct. Plausible is the student does not take into consideration the cooldown rate.

B. INCORRECT: It is desired to maintain a minimum amount of flow to each SG when both are depressurizing in an uncontrolled manner.

Plausible if the student has a misconception on how to deal with two faulted steam generators, by isolating the more faulted or damaged unit.

C. INCORRECT: The direction is to minimize feed flow but maintain flow to both. Plausible if the student has a misconception on how to deal with two faulted steam generators, by isolating the more faulted or damaged unit.

D. CORRECT: Per ECA-2.1, 50 GPM per SG is directed based on the cooldown rate. Examinee will need to calculate a cooldown rate (570°F - 400°F = 170°F in the last 20 minutes)

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 17 of 199 QUESTION # 008 Given the following:

• Unit 2 is shutdown coming out of an outage

• The RCS is aligned to RHR with RCS temperature at 275°F and Stable

• The crew is testing 2CS-3124, Main Feed Isolation Valve, for PMT

• While 2CS-3124 was in the open position, the pneumatic operating supply was lost to the valve What is the final position of 2CS-3124 and the impact on Technical Specifications?

A. Open and transition to the next higher MODE will NOT be allowed.

B. Open and transition to the next higher MODE will be allowed.

C. Closed and transition to the next higher MODE will NOT be allowed.

D. Closed and transition to the next higher MODE will be allowed.

ANSWER:

B.

REFERENCE:

DBD-03, Condensate and Feedwater System, Rev 17 TS 3.7.3 Plant Systems, Feedwater Isolation Valves (MFIVs), Main Feedwater Regulating Valves (MFRVs), and MFRV Bypass Valves, Rev Amendment 241/245 TS B 3.7.7, Bases,Plant Systems, Feedwater Isolation Valves (MFIVs),

Main Feedwater Regulating Valves (MFRVs), and MFRV Bypass Valves, Rev Amendment 241/245 NEW HIGHER K/A: 054G2.2.36: Loss of Main Feedwater: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operation.

LEARNING OBJECTIVE:

DESCRIBE the function and/or purpose, design bases, and operating characteristics of the Feedwater system and major components.

(052.05.LP0128.001)

IDENTIFY and DISCUSS the Technical Specifications associated with Main Feedwater System components, parameters, and operation to include:

Limiting Conditions for Operation (LCO)

LCO applicability (052.05.LP0128.009)

EXPLANATION:

A. CORRECT: Correct position. MFIVs use a pneumatic supply to open and close so valves fail in whatever position they are in at the time of the loss. LCO 3.7.3 MFIVs, MFRVs and MFRV Bypasss is

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 18 of 199 applicable in MODES 1, 2, and 3 with Transition into MODE 3 from MODE 4 not allowed until the LCO is met.

B. INCORRECT: Correct position. LCO 3.7.3 MFIVs, MFRVs and MFRV Bypasss is applicable in MODES 1, 2, and 3 with Transition out of MODE 4 not allowed. Plausible if the student does not know the modes of applicability, or misidentifies the current plant mode.

C. INCORRECT: Wrong position. For the given conditions, transition to the next higher mode is not allowed. Plausible because the student may believe the valve fails to its required position which is closed, like a containment isolation valves and needs to determine the current mode and the mode or applicability for the current event.

D. INCORRECT: Wrong position. LCO 3.7.3 MFIVs, MFRVs and MFRV Bypasss is applicable in MODES 1, 2, and 3 with Transition out of MODE 4 is not allowed. Plausible because the student may believe the valve fails to its required position which is closed, like a containment isolation valves, and needs to determine the current mode and the mode or applicability for the current event.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 19 of 199 QUESTION # 009 Given the following:

• A Station Blackout has occurred on Unit 2

• The crew has worked through the EOP network to successfully restore an EDG locally

• A Component Cooling Water (CCW) Pump was restarted when power was restored on Unit 2 causing a pressure surge

• CC-721C, HX-12C CCW Heat Exchanger Relief Valve, has lifted and will not reseat What indications will be seen for this event?

A. Continuously lowering level in Unit 2 CCW Surge Tank ONLY.

B. Lowering level in Unit 2 CCW Surge Tank and rising level in Unit 1 CCW Surge Tank.

C. Continuously lowering level in Unit 2 CCW Surge Tank and a rising level in the Waste Holdup Tank (WHUT).

D. Level in Unit 2 CCW Surge Tank will NOT change.

ANSWER:

D.

REFERENCE:

110E018 Sh. 3, P&ID Auxiliary Coolant System, Rev 43 NEW HIGHER K/A: 055G2.2.4: Station Blackout: Ability to explain the variations in control board layouts, systems, instrumentation, and procedural actions between units at a facility.

Learning Objective:

DESCRIBE the function and/or purpose, design bases, physical locations and operating characteristics of the Component Cooling Water System and major components. (051.06.LP0084.001)

EXPLANATION:

A. INCORRECT: This is what happens for some CCW system relief valves that dump to the floor or inside containment. Plausible because of the number of relief valve locations and relief paths in the CCW system.

B. INCORRECT: This is the opposite of what happens if a relief lifts on the Unit 1 CCW Heat Exchangers. Plausible because of the number of relief valve locations and relief paths in the CCW system.

C. INCORRECT: Relief is supplied by Unit 2 CCW and some reliefs dump outside the system to the WHUT. Plausible because of the number of relief valve locations and relief paths in the CCW system.

D. CORRECT: Relief valve is supplied by Unit 2 CCW system and relieves back to the Unit 2 CCW Surge tank.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 20 of 199 QUESTION # 010 Considering the Safety Injection Load Sequencer operation during an accident with a Loss of Offsite Power, which of the following describes the reason(s) behind the SI Load Sequencer order and time to initiation of power to the various loads?

A. The considerations deal ONLY with maintaining the diesel generator frequency and voltage within tolerance.

B. The considerations deal ONLY with ensuring the starting of the various engineered safety features are within the required safety analysis response time values.

C. The required response time values of the various emergency safety features are based on the accident analysis of the plant for the design basis accident AND the diesel generator capability.

D. In order to ensure that the diesel generator's speed will not decrease below 95% of nominal value, the largest loads are started first when the diesel generator can best handle the starting currents.

ANSWER:

C.

REFERENCE:

DBD-24, ESF (Safeguards) Actuation System Design basis Doc, Section 2.2.5, Pg. 37, Rev 7 BANK: 2007 PBNP NRC Exam FUNDAMENTAL K/A: 056AK3.01: Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Order and time to initiation of power for the load sequencer.

LEARNING OBJECTIVE:

Describe the automatic functions associated with the ESFAS and its major components. Description should include ...protection afforded by the system. (LP0486.14)

EXPLANATION:

A. INCORRECT: Incorrect, this is only one half of the reason.

B. INCORRECT: Incorrect, this is only one half of the reason.

C. CORRECT: This is the correct answer.

D. INCORRECT: Diesel generator's speed must be maintained and starting currents must be allowed to decay between subsequent equipment starts making this a potential reason.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 21 of 199 QUESTION # 011 Given the following:

• A reactor startup on Unit 1 is in progress

• Reactor power is in the intermediate range and stable when the following occurred

• Most RED bistable status lights are LIT

• The following annunciators are LIT:

o RED INVERTER TROUBLE o UNIT 1 INSTRUMENT BUS UNDER/OVER VOLTAGE What actions are required for this condition?

A. Verify reactor trip AND isolate the Red Instrument Bus.

B. Verify reactor trip AND restore power to the Red Instrument Bus from DYOA.

C. Commence a reactor shutdown to insert all control and shutdown banks AND isolate the Red Instrument Bus.

D. Commence a reactor shutdown to insert all control and shutdown banks AND restore power to the Red Instrument Bus from DYOA.

ANSWER:

B.

REFERENCE:

ARP 2C20 A 4-1, Instrument Bus Under/Over Voltage, Rev 0 ARP 2C20 B 3-8, Red Inverter Trouble, Rev 3 AOP-0.2, Loss of Safety Related Instrument Buses, Attachment A, Rev 5 OR-01, Important Systems, Signals and Setpoints, Rev 0 BANK: 2004 Ginna NRC Exam HIGHER K/A: 057AK3.01: Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: Actions contained in EOP for loss of vital ac electrical instrument bus.

LEARNING OBJECTIVE:

Given access to the site specific simulator or specific plant conditions, RESPOND to the following:

Loss of a DC bus Loss of an Instrumentation Bus (055.03.LP3456.002)

EXPLANATION:

A. INCORRECT: Based on the plant conditions, the reactor will trip, and must be verified by procedure, the second part of the answer is not correct, as the Red Instrument Bus will be powered from an alternate power source.

B. CORRECT: Based on the plant conditions, the reactor will trip, and must be verified by procedure, and then power the instrument bus from an alternate power source.

C. INCORRECT: The reactor will trip; a reactor shutdown will not be necessary with this fault. The second part of the answer is not correct,

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 22 of 199 as the Red Instrument Bus will be powered from an alternate power source.

D. INCORRECT: The reactor will trip; a reactor shutdown will not be necessary with this fault. The second part of the answer is correct; the instrument bus will be powered from an alternate power supply.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 23 of 199 QUESTION # 012 Given the following:

• During a maintenance period the following DC system alignment exists:

o DC Bus D02 is aligned to D09, Swing Battery Charger and D305 Swing Battery o D09, Swing Battery Charger develops a ground and the AC power breaker on D09 Trips Which of the following methods can be used to restore D02 to an acceptable alignment?

A. Crosstie DC Bus D02 to DC Bus D04.

B. Align 2D-205, Battery and 2D-207, Battery Charger to D-02.

C. Restore D08, Battery Charger and D06, Battery to DC Bus D02.

D. Align Swing Battery Charger D109 to DC Bus D02 and D305 Battery.

ANSWER:

C.

REFERENCE:

AOP-0.0, Vital DC System Malfunction, step 5, Rev 34 0-SOP-DC-002, 125 VDC System Bus D-02 & Components, Rev 20 BANK FUNDAMENTAL K/A: 058AA1.01: Ability to operate and / or monitor the following as they apply to the Loss of DC Power: Cross-tie of the affected dc bus with the alternate supply.

LEARNING OBJECTIVE:

STATE why train cross-connection is not allowed. (054.03.LP0121.010)

EXPLANATION:

A. INCORRECT: System interlocks will prevent this lineup. Plausible if the student does not understand the interlock and interconnections of D02 and D04.

B. INCORRECT: System interlocks will prevent this lineup. Plausible is the student does not understand how the system Kirk Key interlocks function.

C. CORRECT:This is the method called out in the Abnormal Operating Procedure, and is achievable based on system design and interlocks.

D. INCORRECT: System interlocks will prevent this lineup. Plausible if the student does not understand the interlock and connections of D301A.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 24 of 199 QUESTION # 013 Given the following:

• Unit 1 was operating at 100% when the unit tripped inadvertently

• Immediate actions are complete

• The following indications are observed:

o 1HX-15A Cont Ventilation SW Outlet Flow 580 gpm o 1HX-15B Cont Ventilation SW Outlet Flow 400 gpm o 1HX-15C Cont Ventilation SW Outlet Flow 590 gpm o 1HX-15D Cont Ventilation SW Outlet Flow 575 gpm

• C01 C01B 1-4, UNIT 1 CONTAINMENT SUMP A LEVEL HIGH annunciator is LIT

• C01 C01B 2-3, UNIT 1 CONTAINMENT RECIRC COOLERS WATER FLOW LOW annunciator is NOT lit What actions are required to respond to the above indications?

A. Isolate the North Service Water Header per AOP 9A, Service Water System Malfunction.

B. The crew will have the Primary Auxiliary Building watch isolate affected component(s) per AOP 9A, Service Water System Malfunction.

C. The crew will verify the service water system alignment per EOP 0, Reactor Trip or Safety Injection, Attachment A, Automatic Action Verification.

D. Manually shut 1SW-2907/2908, Containment Ventilation Coolers Outlet Emergency Flow Control Valves, per Containment Sump A high level alarm response.

ANSWER:

B.

REFERENCE:

AOP-9A, Service Water System Malfunction, Rev 32 M-207 Sh 4, P&ID Service Water System, Rev 30 LP0086, Service Water Lesson Plan Powerpoint, Rev 22 BANK: 2009 PBNP NRC Exam HIGHER K/A: 062AA1.07: Ability to operate and / or monitor the following as they apply to the Loss of Nuclear Service Water (SWS): Flow rates to the components and systems that are serviced by the SWS; interactions among the components.

LEARNING OBJECTIVE:

IDENTIFY and DESCRIBE the Control Room controls, alarms, and indications associated with the Service Water System, including:

Location and function of component and/or system operating controls and control stations Alarming locations and response to major system and component alarm Plant, system, and component conditions or permissives required for

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 25 of 199 Control Room operation Setpoints associated with major system alarms and/or interlocks (051.06.LP0086.011)

DESCRIBE the function and/or purpose, design bases, and operating characteristics of the Service Water system and major components.

(051.06.LP0086.001)

EXPLANATION:

A. INCORRECT: The leaking Service Water cooler can be isolated per AOP-9A, Isolation of the North Header is not required. Plausible is the student determines that isolation of the north header is needed to mitigate the event.

B. CORRECT: The crew will respond to the trip with EOP-0.1 and the leak with AOP-9A C. INCORRECT: Attachment A of EOP-0 is not applicable without an SI, transition would be from the immediate actions to EOP-0.1. Plausible if the student is unable to deduce an SI has not happened given the current conditions.

D. INCORRECT: These valves are currently shut since no SI occurred, based on the Containment recirc cooler flow low alarm NOT lit.

Plausible because the student must deduce that an SI did not happen, and these valves control flow to the accident fans.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 26 of 199 QUESTION # 014 Given the following:

• Both Units have been tripped due to a Loss of Instrument Air

• The instrument air headers are depressurized

• A Main Feedwater isolation signal has actuated How will this Loss of Instrument Air impact the Main Feedwater Isolation Valves and Main feed Regulating Valves?

MFIVs MFRVs A. Open Shut B. Open Open C. Shut Shut D. Shut Open ANSWER:

C.

REFERENCE:

M-209 Sh 5, P&ID Instrument Air, Rev 38 M-202 Sh 2, P&ID Feedwater System, Rev 55 AOP-5B, Loss of Instrument Air, Rev 41 FUNDAMENTAL NEW K/A: 065AA2.07: Ability to determine and interpret the following as they apply to the Loss of Instrument Air: Whether backup nitrogen supply is controlling valve position.

LEARNING OBJECTIVE:

DESCRIBE the function and/or purpose, design bases, and operating characteristics of the Engineered Safety Features Actuation System and its major components/subsystems to include:

a. Containment Isolation

b. Containment ventilation Isolation

c. Containment Spray Actuation

d. Steam Line Isolation

e. Feedline Isolation

f. Safety Injection (053.06.LP0486.001)

EXPLANATION:

A. INCORRECT: The second part is correct. Plausible if the student has a misconception about the MFIV not recalling it is a fail as is valve with a nitrogen backup, and that the MFRV is a fail close on loss of IA valve.

B. INCORRECT: both parts are incorrect. Plausible if the student has a misconception that the MFIV and MRFV are fail as is valves with no back up supplies.

C. CORRECT: The MFIV is a fail as is valve with a nitrogen backup, the loss of instrument air will not impact the operation of this valve for

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 27 of 199 feedwater isolation. The MFRV is a fail closed valve on a loss of instrument air.

D. INCORRECT: The first part is correct. Plausible if the student has the misconception of the MFIV having a nitrogen backup, automatic isolation of the MFRV would not be required, due to the MFIV stopping flow through the MFRV when is goes shut

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 28 of 199 QUESTION # 015 Given the following:

• A LOCA outside containment has resulted in RCS Subcooling dropping to 0°F

• The operating crew is performing the actions of ECA-1.2, LOCA Outside Containment

• Actions are being taken to identify and isolate the leak In accordance with ECA-1.2...

(1) which of the following valves will be sequentially SHUT and then OPENED for leak identification AND (2) what would be the indication used to confirm the LOCA has been isolated?

A. (1) 1SI-852A and B, RHR Reactor Vessel Injection (2) RCS pressure rising.

B. (1) 1SI-852A and B, RHR Reactor Vessel Injection (2) Pressurizer level rising.

C. (1) 1SI-878A and C, SI Reactor Vessel Injection (2) RCS pressure rising.

D. (1) 1SI-878A and C, SI Reactor Vessel Injection (2) Pressurizer level rising.

ANSWER:

A.

REFERENCE:

ECA-1.2, LCOA Outside Containment, Rev 24 MODIFIED: 2010 Sequoyah NRC Exam FUNDAMENTAL K/A: W/E04EA2.2: Ability to determine and interpret the following as they apply to the (LOCA Outside Containment): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

LEARNING OBJECTIVE:

Given appropriate conditions/parameters and access to the site specific Simulator, IMPLEMENT the following procedures for the specified conditions:

a. ECA-1.1 to respond to a loss of Containment Sump Recirculation

b. ECA-1.2 to respond to an intersystem LOCA

c. ECA-1.3 to respond to containment sump blockage

d. ECA-2.1 to respond to both Steam Generators being faulted (031.02.LP0465.008)

EXPLANATION:

A. CORRECT: Per ECA-1.2, this is a leak isolation strategy and a method to determine if the leak was isolated.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 29 of 199 B. INCORRECT: Per ECA-1.2 this is a correct leak isolation strategy, but he method to determine the leak was isolation is wrong. Plausible because this indication will show that the leak is isolated but it is not the required method per ECA-1.2 C. INCORRECT: Per ECA-1.2, this is not a correct leak isolation strategy, and the method to determine if the leak was isolated is correct. These valves are normally shut, and do not get an open signal. Plausible because the student may confuse this with SI cold leg injection, which is an isolation strategy per ECA-1.2.

D. INCORRECT: Per ECA-1.2, this is not a correct leak isolation strategy, and the method to determine the leak was isolated is wrong. These valves are normally shut, and do not get an open signal. Plausible because the student may confuse this with SI cold leg injection, which is an isolation strategy per ECA-1.2 and this leak isolation indication will show that the leak is isolated but it is not the required method per ECA-1.2

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 30 of 199 Original Question Given the following plant conditions:

• Following a reactor trip, abnormal radiation was noted in the Aux.

Building due to a loss of RCS inventory outside containment.

Which ONE of the following identifies a required action and the subsequent check used to determine whether or not the leak is isolated in accordance with ECA-1.2, "LOCA Outside Containment?"

A. Isolate SI pump Cold Leg Injection; Pressurizer level rising B. Isolate SI pump Cold Leg Injection; RCS pressure rising C. Isolate RHR Cold Leg Injection; Pressurizer level rising D. Isolate RHR Cold Leg Injection; RCS pressure rising

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 31 of 199 QUESTION # 016 Given the following:

• Unit 1 operators are attempting to place ECCS on Containment Sump recirculation following a LOCA Which of the following would prevent ANY sump recirculation flow from being established?

A. RCS pressure is greater than 475 PSIG B. 1SI-896B, B SI Pump Suction from RWST isolation valve OPEN C. BOTH 1SI-897A and B, SI Test Line Return isolation AOVs OPEN D. 1SI-857A, A RHR Heat Exchanger Outlet to SI pump suction valve SHUT ANSWER:

C.

REFERENCE:

EOP-1.4, Transfer to Containment Sump Recirculation High Head Injection, Rev 29 EOP-1.3, Transfer to Containment Sump Recirculation Low Head Injection, Rev 52 BG EOP-1.4, Background Document Transfer to Containment Sump Recirculation High Head Injection, Rev 34 BG EOP-1.3, Background Document Transfer to Containment Sump Recirculation Low Head Injection, Rev 17 BANK: 2007 PNBP NRC Exam HIGHER K/A:: W/E11EK2.1: Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following: Components and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes and automatic and manual features.

LEARNING OBJECTIVE:

Describe the interlocks associated with the Safety Injection System and its major components. (051.03.LP0066.004)

EXPLANATION:

A. INCORRECT: The purpose of both EOP-1.3 and EOP-1.4 is to establish sump recirculation, one for high pressure and the other for low (greater than or less than 325 psig). Plausible if the student does not recall the fact that there are two procedures, also plausible given the first step in the procedure after identification of the system pressure is to secure the RHR pumps.

B. INCORRECT: This will prevent placing train B of the ECCS system in sump recirculation, but will not prevent train A. Plausible if the student does not understand the structure of the procedure to attempt one train and then the other.

C. CORRECT: This will prevent sump recirculation, at least one must be closed in both the High Head and Low Head procedures.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 32 of 199 D. INCORRECT: This will prevent placing train A of the ECCS system in sump recirculation, but will not prevent train B. Plausible if the student does not understand the structure of the procedure to attempt one train and then the other.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 33 of 199 QUESTION # 017 Given the following:

• Unit 1 was at Rated Thermal Power when a large break LOCA occurred

• The ECCS system is functioning as designed

• The crew has just transitioned to EOP-1, Loss of Reactor or Secondary Coolant

• Current Plant Parameters:

o Wide Range RCS Pressure 25 psig Stable o Core Exit Thermocouples 237°F Stable o RCS Cold Leg Loop Temperatures 150°F Stable o Steam Generator Pressures 400 psig Stable o Steam Generator 'A' Narrow 2% Slowly Rising Range Level o Steam Generator 'B' Narrow 4% Slowly Rising Range Level o Total AFW Flow to Both 230 gpm Stable Steam Generators o Containment Pressure 25 psig Slowly Lowering o Pressurizer Level 0%

o NR RVLIS 25 feet Slowly Lowering o RWST Level 50% Slowly Lowering What is the current mechanism by which core decay heat is being removed?

A. AFW flow B. RCS break flow C. Two phase flow through the core D. Heat transfer to Component Cooling from RHR ANSWER:

B.

REFERENCE:

BG CSP-H.1, Background Document Response to Loss of Secondary Heat Sink Steam Tables, Ref 27 EOP-1.3, Transfer to Containment Sump Recirculation - Low head Injection, Rev 52 BANK: 2011 Kewaunee NRC Exam HIGHER K/A: W/E05EK2.2: Knowledge of the interrelations between the (Loss of Secondary Heat Sink) 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.

LEARNING OBJECTIVE:

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 34 of 199 PREDICT the effects of a loss of Reactor Coolant Flow (043.02LP2461.009)

Given appropriate conditions/parameters or access to the Site Specific Simulator, ANALYZE the following:

Relationships between Reactor Coolant System, Safety Injection Flow and Break Flow Indications when determining Reactor Coolant System conditions (Subcooled, Saturated, Superheated) (031.02.LP0435.002)

Given appropriate conditions/parameters or access to the Site Specific Simulator, RECOGNIZE the following conditions/events:

Significance of a loss of a Steam Generator on heat removal capability (031.02.LP0435.004)

EXPLANATION:

A. INCORRECT: During a large break LOCA the steam generators become thermally disconnected from the primary. Plausible because this is the normal heat sink for the primary.

B. CORRECT: RCS break flow is the method of heat removal during a large break LOCA C. INCORRECT: Given the pressure of the RCS, and the temperature of the CET,s subcooling exists. Based on the subcooling, the reactor coolant pumps will be off. Plausible because this is a method of cooling earlier in the event, or during a small break LOCA.

D. INCORRECT: With the RWST at the current level, the RHR system is injection mode. Transfer to containment sump recirculation will not happen until 34%. Plausible because at the current RWST level, the crew should have transitioned to EOP -1.3, and lining up for containment sump recirculation.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 35 of 199 QUESTION # 018 Given the following:

• Unit 2 is in MODE 1

• Grid disturbances have required the crew to lower turbine generator output to 470 MWE

• Generator Hydrogen pressure is 60 psig Which of the following is the maximum VAR loading per the capability curve the generator may carry without exceeding limits?

(See provided reference)

A. 235 Mvars overexcited B. 295 Mvars overexcited C. 320 Mvars overexcited D. 330 Mvars overexcited ANSWER:

B.

REFERENCE:

OP 2A, Normal Power Operations, Rev 6 Proposed reference to be provided to the applicants during examination:

OP-2A Unit 2, Figure 3 (page 56)

MODIFIED HIGHER K/A:: 077AK1.02: Knowledge of the operational implications of the following concepts as they apply to Generator Voltage and Electric Grid Disturbances: Over-excitation.

LEARNING OBJECTIVE:

Given appropriate system/equipment conditions and indications, DESCRIBE and DEMONSTRATE the ability to perform/control the following evolutions:

a. Increase power as Xenon changes

b. Control Delta Flux

c. Determine Reactor Thermal output

d. Control MVARS as directed by the System Control Supervisor

e. Predict Xenon transients

f. Estimate Rod position at power and during power swings such that Delta Flux remains within Technical Specifications limits

g. Transfer 4 kV non-vital busses between transformers (055.01.LP0258.002)

EXPLANATION:

A. INCORRECT: 235 Mvars overexcited is the result of using Figure 3 with the intersection being at 470 MWE and 45 psig hydrogen.

Plausible if the student uses the wrong curve.

B. CORRECT: 295 Mvars overexcited is the result of using Figure 3 with the intersection being at 470 MWE and 60 psig hydrogen, so a value of 295 was selected due to graph readability.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 36 of 199 C. INCORRECT: 320 Mvars overexcited is the result of using Figure 3 with the intersection at 470 MWE and 75 psig hydrogen, but using the lower or underexcited portion of the graph. Plausible if the student uses the wrong half of Figure 3.

D. INCORRECT: 330 Mvars overexcited is the result of using Figure 3 with the intersection being at 470 MWE and 75 psig hydrogen, or the normal curve on the graph. Plausible if the student has a misconception of how to use the Figure 3.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 37 of 199 Original Question:

Given the following:

• Unit 2 is in MODE 1

• Turbine generator output is 450 MWE

• Generator Hydrogen pressure is 60 psig Which if the following is the maximum VAR loading per the capability curve the generator may carry without exceeding limits?

A. 240 Mvars overexcited B. 270 Mvars overexcited C. 310 Mvars overexcited D. 340 Mvars overexcited

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 38 of 199 QUESTION # 019 Given the following:

Time = 1000

• Reactor Power is 65% and Stable

• TAVG - TREF deviation is 0°F and Stable

• PZR level is 37% and Stable

• Rods are in AUTO

• Control Bank D step counters are at 165 steps Time = 1002

• Reactor Power is approximately 66% and rising (no load change is in progress)

• TAVG - TREF deviation is approximately +2°F and Rising

• PZR level 38% and Rising

• PZR spray valves have throttled open

• Control Bank D step counters are at 174 steps and stepping out at 8 steps per minute Which of the following describes:

(1) the event in progress AND (2) the FIRST action that must be performed per AOP-6C, Uncontrolled Motion of RCCA(s)?

A. (1) First stage pressure channel high failure.

(2) Trip the Reactor and go to EOP-0, Reactor Trip or Safety Injection.

B. (1) First stage pressure channel high failure.

(2) Place the Rod Control Bank Selector Switch to MANUAL.

C. (1) Continuous Spurious Control Bank Withdrawal.

(2) Trip the Reactor and go to EOP-0, Reactor Trip or Safety Injection.

D. (1) Continuous Spurious Control Bank Withdrawal.

(2) Place the Rod Control Bank Selector Switch to MANUAL.

ANSWER:

D.

REFERENCE:

AOP-6C, Uncontrolled Motion of RCCA(s), Rev 17 BANK: 2011 Harris NRC Exam HIGHER K/A: 001AK2.01: Knowledge of the interrelations between the Continuous Rod Withdrawal and the following: Rod bank step counters.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 39 of 199 LEARNING OBJECTIVE:

DESCRIBE the response of the Rod Control System to failures and/or malfunctions of the following:

a. Nuclear Instrumentation (NIS) Power Range Channel

b. Reactor Coolant Bypass Loop RTD failure

c. Turbine First stage Pressure Transmitter

d. Urgent Failure Alarm (055.03.LP2441.002)

EXPLANATION:

A. INCORRECT: Plausible because this failure will cause rods to move out both for temperature mismatch and for power rate mismatch. The action taken is plausible because it is a required action if rod motion continues after the rod control bank selector switch has been taken to manual.

B. INCORRECT: Plausible because this failure will cause rods to move out both for temperature mismatch and for power rate mismatch. The action taken is the correct action for uncontrolled control rod motion.

C. INCORRECT: Plausible since this is the correct accident. The action taken is plausible because it is a required action if rod motion continues after the rod control bank selector switch has been taken to manual.

D. CORRECT: During a control rod withdrawal TAVG will rise due to the addition of positive reactivity. The stem of the question states the initial control rod position is 165 steps and final position is 174 steps.

Given this a Control Rod withdrawal is taking place. The stem also states there is a +2°F TAVG - TREF deviation and no load change in progress. With Control rods in auto, they should be stepping in.

Immediate actions per AOP-6C are to check rod motion required, which it is, but in the opposite direction, then place rods in Manual and verify that rod motion is stopped.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 40 of 199 QUESTION # 020 Given the following:

• Unit 2 reactor tripped

• Trip and Bypass breakers are open

• Neutron flux is lowering

• One control rod is stuck at 100 steps with the rod bottom light not lit

• The crew is responding to the stuck rod in EOP 0.1, Reactor Trip

Response

Is emergency boration required for the stuck rod and if so from where and why?

A. Boration is not required at this time for shutdown margin.

No immediate action is required since the core is designed for adequate shutdown margin in the current condition.

B. 2825 gallons of boration from a HUT is required.

The shutdown reactivity margin must be made up through emergency boration to account for the reactivity worth of the stuck rod.

C. Boration is not required at this time for shutdown margin.

The emergency core cooling system adds shutdown reactivity, so no consequential damage to the core occurs.

D. 2825 gallons of boration from an RWST is required.

The shutdown reactivity margin must be made up through emergency boration to account for the reactivity worth of the stuck rod.

ANSWER:

A.

REFERENCE:

EOP-0.1 Response to Reactor Trip Step 3, Rev 41 BG EOP-0.1 Background Document Response to Reactor Trip, Rev 30 NEW FUNDAMENTAL K/A: 005AK3.06: Knowledge of the reasons for the following responses as they apply to Inoperable/Stuck Control Rod: Actions contained in EOP for inoperable/stuck control rod.

LEARNING OBJECTIVE:

List the major actions accomplished by each of the following Emergency Procedures.

a. EOP-0

b. EOP-0.0

c. EOP-0.1

d. EOP-1.1 (031.02.LP0405.003)

EXPLANATION:

A. CORRECT: Plant is analyzed for one stuck rod and still having acceptable shutdown margin.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 41 of 199 B. INCORRECT: Incorrect flowpath for boration and amount is what is used per rod. Plausible because this is the value of boration used if more than one rod is stuck and also this is one of the possible sources for borated water during other plant operations, and this is the reason the boration is performed if there is more than one stuck rod.

C. INCORRECT: No boration is required, and the reason why is incorrect.

The ECCS system will add shutdown reactivity, but the procedure path is from EOP-0 to EOP-0.1, ECCS equipment is not running. Plausible because boration is not required, and the ECCS system will add negative reactivity.

D. INCORRECT: Incorrect flowpath for boration and amount is what is used per rod. Plausible because this is the value of boration used if more than one rod is stuck and also this is one of the possible sources for borated water during other plant operations, and this is the reason the boration is performed if there is more than one stuck rod.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 42 of 199 QUESTION # 021 Given the following:

• Unit 1 is at Rated Thermal Power

• Charging pumps are in manual due to a Pressurizer level control malfunction

• Initial plant conditions are o Pressurizer level is stable o 1TI-126, Regenerative Heat Exchanger Charging Outlet temperature is 499°F o 1TI-127, Regenerative Heat Exchanger Letdown Outlet temperature is 262°F o 1PI-135, Letdown Line Pressure is 257 psig

• A Chemical and Volume Control System malfunction occurs resulting in the following conditions:

o Pressurizer level is lowering slowly o 1TI-126, Regenerative Heat Exchanger Charging Outlet temperature is 539°F o 1TI-127, Regenerative Heat Exchanger Letdown Outlet temperature is 376°F o 1PI-135, Letdown Line Pressure is 257 psig Which of the following malfunctions would explain the observed change in indications?

A. Loss of air to 1CV-142, Charging Header Flow Control Valve B. Loss of air to 1CV-135, Letdown Back Pressure Control Valve C. A Letdown line leak upstream of the Regenerative Heat Exchanger D. A Charging Header leak upstream of the Regenerative Heat Exchanger ANSWER:

D.

REFERENCE:

684J741 Sh 2, P&ID Chemical and Volume Control, Rev 73 684J741 Sh 3, P&ID Chemical and Volume Control, Rev 17 BANK: 2013 Comanche Peak NRC Exam HIGHER K/A: 028AA1.06: Ability to operate and / or monitor the following as they apply to the Pressurizer Level Control Malfunctions: Checking of RCS leaks.

LEARNING OBJECTIVE:

DESCRIBE the plant and operator(s) response to the following conditions:

a. Failure of Pressurizer Pressure and/or Level Control system

b. Reactor Coolant System leak

c. Reactor Coolant Pump malfunction

d. Steam Generator Tube leak

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 43 of 199

e. Loose Parts Monitoring System alarm

f. Rapid Power Reduction. (055.03.LP2438.001)

EXPLANATION:

A. INCORRECT: If air was lost to the charging header flow control valve the valve will fail open, causing more charging flow to go through the regenerative heat exchanger, causing temperatures of charging outlet and letdown outlet to lower, it will also cause a rise in pressurizer level.

Plausible because if the student has a misconception of the failed position and believes it will fail closed, that will cause the same effects as a leak.

B. INCORRECT: If air was lost to the letdown back pressure control valve, it would fail open causing more letdown flow. Temperatures for the charging outlet and letdown outlet will both rise, but only slightly, and pressurizer level will remain constant due to the flow control orifice that is in the letdown line. Plausible because this will cause the same type of effect on temperature indications, but not to the same degree.

C. INCORRECT: A leak at this location will cause the pressurizer level to lower, but the charging outlet temperature will lower, not rise as indicated. Plausible because if the student has a misconception on the flow path of the regenerative heat exchanger and the effects of a letdown leak.

D. CORRECT: A leak here would lower the charging flow to the regenerative heat exchanger. This reduced flow would cause the both the letdown outlet temperature and charging outlet temperature to rise as well as pressurizer level to lower.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 44 of 199 QUESTION # 022 Given the following:

• Unit 1 is performing a power ascension following a refueling outage

• During the power ascension the log readings for N-35 and N-36 IR NIs have trended as follows:

N-35 N-36 0800 (2/1) 5.1E-10 2.1E-9 1600 (2/1) 2.1E-8 8.1E-8 0000 (2/2) 6.9E-7 2.8E-6 0800 (2/2) 1.0E-5 7.9E-5 Which of the following indicate when control board readings for N-35 and N-36 first exceeded the maximum allowable channel deviation?

(A picture representation of the meter readings is on the following page)

A. 0800 (2/1)

B. 1600 (2/1)

C. 0000 (2/2)

D. 0800 (2/2)

ANSWER:

D.

REFERENCE:

PBF-2034 Unit 1 Control Room Log, page 83, Rev 86 MODIFIED: 2010 Sequoyah NRC Exam HIGHER K/A: 033AA2.12: Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation:

Maximum allowable channel disagreement.

LEARNING OBJECTIVE:

DESCRIBE the operation and setpoints of interlocks, Permissives, and Automatic actions associated with the Nuclear instrumentation System

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 45 of 199 EXPLANATION:

Per PBF-2034 Unit 1 Control Room Logs, the maximum allowable deviation between intermediate range channels, is 0.7 decades. The table below is shows the deviation from the values used in the question.

Date/Time N-35 N-36 Log of Log of Channel N-35 N-36 difference 0800 (2/1) 5.10E-10 2.10E-09 9.292 8.678 0.614 1600 (2/1) 2.10E-08 8.10E-08 7.678 7.092 0.586 0000 (2/2) 6.90E-07 2.80E-06 6.161 5.553 0.608 0800 (2/2) 1.00E-05 7.90E-05 5 4.102 0.898 A. INCORRECT: Plausible because the channel deviation is 0.6, and there is a 0.7 of a decade difference using addition/subtraction, in the channel readings B. INCORRECT: Plausible because the channel deviation is .58.

C. INCORRECT: Plausible because the channel deviation is 0.60.

D. CORRECT: See above.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 46 of 199 Original Question:

Given the following:

• Unit 1 is in Mode 1 doing a power ascension following a refueling outage.

• During the power ascension the 1-M-4 Control Board indicators for N-35 and N-36 (1-X1-92-5003A & 1-X1-92-5004A) have trended as follows:

N-35 N-36 2000 (6/18) 8x100 4x101 0800 (6/19) 1x101 6x101 2000 (6/19) 2x101 1x102 Which ONE of the following indicates the status of control board readings for N-35 and N-36?

REFERENCE PROVIDED A. maximum allowable channel deviation exceeded at 2000 (6/18)

B. maximum allowable channel deviation exceeded at 0800 (6/19)

C. maximum allowable channel deviation exceeded at 2000 (6/19)

D. maximum allowable channel deviation not exceeded.

(053.03.LP2416.011)

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 47 of 199

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 48 of 199 QUESTION # 023 Given the following:

• Unit 1 reactor power is 75%

The following plant conditions currently exist:

• A Steam Generator Tube Leak is occurring

• The crew is responding per AOP-3, Steam Generator Tube Leak

• Pressurizer Level is 28% and Very Slowly Lowering

• Letdown has been isolated

• Charging flow is 138 gpm

• RCP Seal Injection supply flow is 9 gpm per pump

• RCP seal return flow is 1.6 gpm per pump

• The leaking Steam Generator Level is 64% and Stable

• RCS Subcooling is 28°F and Stable Based on the information provided above which of the following statements is correct?

A. Charging flow is low for the given conditions and should be adjusted higher.

B. Pressurizer Level is low for these conditions and the reactor should be tripped.

C. RCP seal injection flow is high for these conditions and should be locally adjusted.

D. Steam Generator Level is high for these conditions and should be manually adjusted.

ANSWER:

B.

REFERENCE:

AOP-3, Steam Generator Tube Leak, Step 3 (continuous action statement), Rev 10 OM 3.7, AOP and EOP Procedure Usage for Response to Plant Transients, Rev 23 AOP-1B, Reactor Coolant Pump Malfunction, Rev 22 MODIFIED HIGHER K/A: 037G2.4.1: Steam Generator Tube Leak: Knowledge of EOP entry conditions and immediate action steps.

LEARNING OBJECTIVE:

Given access to the Site Specific Simulator or specific plant conditions, RESPOND to the following events:

a. Reactor Coolant System leakage

b. Reactor Coolant Pump malfunctions

c. Steam Generator Tube leak

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 49 of 199

d. Loose Parts Monitoring System alarm

e. Rapid Power Reduction (055.03.LP2438.004)

EXPLANATION:

A. INCORRECT: Charging is near the maximum rate (140 gpm per OMM 3.7 section 5.0). Plausible if the student does not remember the maximum charging limit, as has the incorrect concept of maximum charging being the maximum possible charging system output, which would be 180 gpm (60 gpm/pump), with the initial condition for the pressurizer level being very slowly lowering, this would lead the student to believe a small adjustment in the charging flow may maintain level.

B. CORRECT: Pressurizer programmed level is 20% at 547°F and 46.1% at 576°F or 47% at 577°F (clipped). Regardless of which value the student utilizes (46.1 or 47%), the current pressurizer level will be lower by greater than 10% (using 46.1% results in 39.57% and using 47% results in 40.25%). Per step 3 of AOP-3, letdown is isolated, and then conditions of step 1 are met, both steps continuous action steps, if the pressurizer is not maintained within 10% of program level and RCS subcooling is greater than 30°F, then a reactor trip and transition to EOP-0 is required.

C. INCORRECT: Seal injection is required to be 6-10 gpm per AOP-1B, Step so no adjustment is needed. Plausible if the student has the misconception that 9 gpm is too high of an injection rate.

D. INCORRECT: Steam generator level is within the allowable range.

Plausible if the student assumes the requirement to maintain the steam generator level during the cooldown requires additional actions.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 50 of 199 Original Question:

Unit 1 reactor power is 75% and steady. The following stable plant conditions currently exist:

• A 17 gpm Steam Generator Tube Leak is occurring

• The Unit 1 operators have stabilized Pressurizer Level at 28%

• Letdown flow rate is 40 gpm

• Charging flow is 44 gpm

• RCP seal return flow is 1.5 gpm per pump

• Steam Generator Level is 64% and stable on the leaking Steam Generator

• RCP Seal Injection supply flow is 8 gpm per pump Based on the information provided above which of the following statements is correct?

A. Charging flow is high for these conditions.

B. Steam Generator Level is low for these conditions C. RCP seal return flow is high for these conditions.

D. Pressurizer Level is low for these conditions.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 51 of 199 QUESTION # 024 Which of the following statements lists the type of release that would be expected during an accidental spill of liquid from the tank listed?

Tank Type of Release from Tank A. Holdup Tank Dissolved fission gases, mostly beta-gamma and alpha radiation B. Waste Holdup Tank Low levels of dissolved fission gases, some delayed neutron and alpha radiation C. Volume Control Tank Dissolved fission gases, mostly delayed neutron and alpha radiation D. Reactor Makeup Tank Low levels of dissolved fission gases, mostly beta-gamma radiation ANSWER:

A.

REFERENCE:

FSAR Section 9.3, Chemical and Volume Control System, Rev 2014 FSAR Table 9.3-5, Reactor Coolant System Equilibrium Activities, Rev 2014 MODIFIED: 2010 Prairie Island NRC Exam HIGHER K/A: 059AK1.01: Knowledge of the operational implications of the following concepts as they apply to Accidental Liquid Radwaste Release:

Types of radiation, their units of intensity and the location of the sources of radiation in a nuclear power plant.

LEARNING OBJECTIVE:

WRITE and EXPLAIN the N16 reaction associated with the Reactor Coolant System during power operations. (051.01.LP0083.008)

DESCRIBE the function and/or purpose, design bases, and operating characteristics of the Chemical Volume and Control System. Description should include design operation of the following specific components:

d. Volume Control Tank (051.02.LP0079.001)

DESCRIBE the function and/or purpose, design bases, and operating characteristics of the Liquid Waste Disposal System and major components. (051.04.LP0063.001)

EXPLANATION A. CORRECT: Hold up Tanks will have water from the Letdown system which would contain the listed types of radiation.

B. INCORRECT: Waste Hold up Tank would not contain water from a source which has delayed neutrons. Plausible is the student has a misconception of the life of delay neutrons and the sources.

C. INCORRECT: Volume Control Tank would not contain water form a source which has delayed neutrons. Plausible is the student has a misconception of the life of delay neutrons and the sources.

D. INCORRECT: Reactor Makeup Water is demineralized water, therefore has no radiation concerns. Plausible if the student has a

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 52 of 199 misconception that reactor makeup water comes from the monitor tanks, which was the original plant design.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 53 of 199 Original Question:

Given the following release types:

• Dissolved fission gases

• Beta-Gamma radiation

• Low level dissolved fission gases

• Delayed neutron and alpha radiation An accidental spill from the would result in a release of .

A. Non-Aerated Sump Tank 1 and 2 B. Waste Holdup Tank 3 and 4 C. Volume Control Tank 1 and 4 D. Reactor Makeup Tank 1 and 2

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 54 of 199 QUESTION # 025 Given the following:

• Unit 2 has tripped

• The crew has entered CSP-H.2, Response to Steam Generator Overpressure, based upon a YELLOW condition on the Heat Sink CSF Status Tree

• SG A pressure is 1150 psig

• SG B pressure is 1050 psig

• SG A Narrow range is 75%

• Instrument Air Header pressure has been lost to Unit 2 Which of the following actions will mitigate the Steam Generator overpressure condition in accordance with CSP-H.2?

A. Raise the SG Blowdown flow rate for the A S/G to lower level B. Open the A SG MSIV Bypass Valve and/or steam 2P-29, TDAFW Pump from the A SG C. Transition to CSP-H.3, Response to Steam Generator High Level to reduce the pressure in the A SG D. Place the Condenser Steam Dump Mode Selector Switch in MANUAL and raise the demand on 2HFC-484, Cond Steam Dump Controller ANSWER:

B.

REFERENCE:

CSP-H.2, Response to Steam Generator Overpressure, Rev 13 HIGHER BANK: 2010 Ginna NRC Exam K/A: W/E13EA2.2: Ability to determine and interpret the following as they apply to Steam Generator Over-Pressure: Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

LEARNING OBJECTIVE:

APPRAISE and PRIORITIZE each operator-initiated recovery technique in its ability to restore the Heat Sink Critical Safety Function.

43.03.LP1998.007)

EXPLANATION:

A. INCORRECT: This will reduce pressure but is not procedurally directed. Plausible because this would work to reduce pressure.

B. CORRECT: without instrument air, these can be opened to reduce pressure and are directed by procedure per step 4.

C. INCORRECT: The transition is not required at this level. Plausible because the level is higher than normal, and is a transition point is the procedure.

D. INCORRECT: Instrument air is not available so this will not work.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 55 of 199 Plausible because this is a procedurally directed step which would e used if instrument air is available.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 56 of 199 QUESTION # 026 Given the following:

• Unit 2 is responding to a LOCA

• The STA has identified high radiation condition in containment

• The OS has you manually actuate Safety Injection and Containment Isolation

• You report the following to the OS:

o RS-SA-10, Radwaste Steam Supply is OPEN o Isolation Panel B 2MS-2083, A SG Sample Containment Isolation Valve Is NOT LIT What actions are you required to take for these conditions?

A. Only shutting 2MS-2083 from Control B. Shut 2MS-2083 from Control, and have the PAB AO shut RS-SA-10 C. Have the PAB AO locally shut both 2MS-2083 and RS-SA-10 D. Shut both 2MS-2083 and RS-SA-10 from Control ANSWER:

B.

REFERENCE:

EOP-0, Reactor Trip or Safety Injection, Unit 2, step 6, Attachment A, Attachment B, Rev 58 NEW FUNDAMENTAL K/A: W/E16G2.1.30 High Containment Radiation: Ability to locate and operate components including local controls.

LEARNING OBJECTIVE:

Given access to the Site Specific Simulator or specific plant conditions, ASSESS response of the Safeguards system to specified accident conditions. Assessments should include conditions requiring actuation and the systems response to each actuation signal.

(031.02.LP0405.031)

EXPLANATION:

A. INCORRECT: 2MS-2083 is operated from the Control Room and must be manually shut. Radwaste Steam could be supplied from Unit 1 but you would have a green light indication instead. Plausible because this is a possible lineup and the student is required to know the Radwaste steam and what they mean.

B. CORRECT: 2MS-2083 is operated from the Control Room and must be manually shut. Radwaste Steam RS-SA-10 is locally operated in the PAB and should be shut locally. Valve indication is the only thing in the control room.

C. INCORRECT: There are containment isolation valves operated in the

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 57 of 199 PAB near the sample room and C-59. Steam RS-SA-10 is locally operated in the PAB and should be shut locally. Valve indication is the only thing in the control room. Plausible due to having indication in the control room, the student may assume there are controls in the control room.

D. INCORRECT: 2MS-2083 is operated from the Control Room and must be manually shut. Plausible because the student may think there are controls for RS-SA-10 in the control room. Valve indication is the only thing in the control room.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 58 of 199 QUESTION # 027 Given the following:

• Unit 1 was operating at Rated Thermal Power when a loss of all off-site power to that unit caused a reactor trip

• Twenty minutes after the trip the following plant conditions exist:

RCS Pressure 2235 psig Stable RCS Hot Leg Temperature 568°F Lowering RCS Cold Leg Temperature 555°F Lowering Core Exit Temperature 580°F Lowering Steam Generator Pressure 1065 psig Lowering Which of the following describes plant conditions?

A. Heat removal IS BEING maintained by Condenser Steam Dumps.

Natural Circulation EXISTS.

B. Heat removal SHOULD BE established by opening the Atmospheric Steam Dumps.

Natural Circulation DOES NOT exist.

C. Heat removal SHOULD BE established by opening the Condenser Steam Dumps.

Natural Circulation DOES NOT exist.

D. Heat removal IS BEING maintained by Atmospheric Steam Dumps.

Natural Circulation EXISTS.

ANSWER:

D.

REFERENCE:

EOP-0.1, Natural Circulation Cooldown, Attchment A, Rev 42 PBN LP0407, Natural Circulation Cooldown, Rev 17 BANK: 2009 Callaway NRC Exam HIGHER K/A: W/E09EK1.3: Knowledge of the operational implications of the following concepts as they apply to the (Natural Circulation Operations):

Annunciators and conditions indicating signals, and remedial actions associated with the (Natural Circulation Operations).

LEARNING OBJECTIVE:

Given access to the Site Specific Simulator, IMPLEMENT the following procedures:

1. EOP-0.2 to cooldown the plant on natural circulation

2. EOP-0.3 to cooldown the RCS when voids are present in the upper head with RVLIS

3. EOP-0.4 to cooldown the RCS when voids are present in the upper head without RVLIS (031.02.LP0407.003)

EXPLANATION:

A. INCORRECT: Condenser steam dumps are not available due to loss

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 59 of 199 of offsite power. Natural circulation does exist. Plausible because the condenser dumps are used during a cooldown normally.

B. INCORRECT: Atmospheric Steam Dumps are the available means to remove decay heat. Natural circulation does exist. Plausible because the atmospheric steam dumps are used, and if the student has a misconception of natural circulation based on the high steam generator pressure.

C. INCORRECT: Condenser steam dumps are not available due to loss of offsite power. Natural circulation does exist. Plausible because the condenser dumps are used during a cooldown normally, and if the student has a misconception of natural circulation based on the high steam generator pressure D. CORRECT: Based on pressure and temperature trends, natural circulation exists and the atmospheric steam dumps will be the method used given the plant conditions.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 60 of 199 QUESTION # 028 Per OP 4B, Reactor Coolant Pump Operation, select the acceptable condition(s) below for starting a Reactor Coolant Pump.

A. The RCS is solid, LTOP is in service, and RCS pressure is 250 psig.

B. The RCS Cold Leg temperature is 200°F and Steam Generator 'A' is 240°F, and 'B' is 245°F.

C. The RCP Lift Pump has been running for 60 seconds, the amber RCP Lift Pressure light is illuminated.

D. The Reactor Coolant Pump was started, run for one minute then stopped. 15 minutes has elapsed since the pump was shutdown.

ANSWER:

B.

REFERENCE:

OP 4B, Reactor Coolant Pump Operation, Rev 57 TS 3.4.6, Reactor Coolant System (RCS), RCS Loops - Mode 4, Rev Amend 246/250 TS 3.4.7, Reactor Coolant System (RCS), RCS Loops - Mode 5, Loops filled, Rev Amend 246/250 TRM 2.2, Pressure Temperature Limits Report, Rev 9 BANK: 2011 PBNP NRC EXAM (Last 2 NRC Exams)

FUNDAMENTAL Correct answer was modified from The RCS Cold Leg temperature is 200°F and the associated Steam Generator is at 240°F, due to change in the reference material.

K/A: 003K6.14: Knowledge of the effect of a loss or malfunction will have on the RCPs: Starting requirements (Imp 2.6/2.9)

LEARNING OBJECTIVE:

Describe the procedures which govern operation of the Reactor Coolant Pump System. Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed or Auxiliary Operators. State the major actions accomplished by each of the following Critical Safety Procedures: (051.01.LP0125.005)

EXPLANATION:

A. INCORRECT: If starting a Reactor Coolant Pump with the RCS solid and LTOP in service then RCS Pressure must be 275 to 325 psig (Initial Condition 4.8). Plausible if the student does not recall the pressure requirements for LTOP.

B. CORRECT: No Reactor Coolant Pump shall be started with one or more RCS cold leg temperatures < Low Temperature Overpressure Protection (LTOP) arming temperature specified in the Pressure and Temperature Limits Report (PTLR) unless the secondary side water temperature of each Steam Generator is < 50°F above each of the RCS cold leg temperatures. (TS 3.4.6, TS 3.4.7)

C. INCORRECT: This Lift Pump is required to run a minimum of two

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 61 of 199 minutes prior to the start of a Reactor Coolant Pump. (Step 5.1.9).

Plausible is the student does not recall the time requirement of the Lift Oil Pump.

D. INCORRECT: Within any 2-hour period, the number of starts should be limited to a maximum of 3 with a minimum idle period of 30 minutes before each restart. (P&L 3.5.3). Plausible if the student does not recall the starting frequency precautions.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 62 of 199 QUESTION # 029 Given the following:

• Unit 1 is operating at Rated Thermal Power at middle of life (MOL)

• The standby CVCS HOH Mixed Bed was last placed in service 45 days ago

• The isolation valve for the standby HOH Mixed Bed has vibrated partially open What effect will this malfunction have on the operation of the plant?

A. Reactor power will rise as RCS temperature rises.

B. Reactor power will rise as RCS temperature lowers.

C. Reactor power will lower as RCS temperature lowers.

D. Reactor power will lower as RCS temperature rises.

ANSWER:

D.

REFERENCE:

OP 5D Part 7, CVCS DEMINERALIZER OPERATION, Rev 6 NEW HIGHER K/A: 004K6.37: Knowledge of the effect of a loss or malfunction on the following CVCS components: Boron loading of demineralizer resin.

LEARNING OBJECTIVE:

Describe the conditions that can cause a change in boron concentration when operating demineralizers, including reactivity concerns. (objective 19 from GFE lesson plan)

RELATE the effects of boron injection and/or dilution Reactor Core reactivity and resulting power deviation, including effects of changing demineralizer alignments. (051.02.LP0079.006)

EXPLANATION:

A. INCORRECT: Both changes are opposite of actual. Plausible if the student has a misconception of the amount of boron that will be in the demineralizer.

B. INCORRECT: Correct temperature change. Plausible if the student has a misconception on the purpose of the demineralizer, and assumes the demin has not been flushed in.

C. CORRECT: HOH Demineralizer has not been in service for a long period of time and is at a higher boron concentration than the unit.

With the leaking valve, more boron will be added to the CVCS system causing RCS temperature to lower and also making reactor power lower.

D. INCORRECT: Correct power change. Plausible if the student equates the power change directly to RCS temperature given a program TAVE.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 63 of 199 QUESTION # 030 Given the following:

• Unit 1 is cooling down on RHR

• RCS temperature is 295 °F and Slowly Lowering

• RCS pressure is 310 psig and Lowering

• Pressurizer level is 80% and Slowly Lowering

• A RCP is operating

• Sump A level is 29% and Rising Which of the following would cause these indications and what action will be taken?

A. 1CV-203, Letdown Orifice Outlet Relief valve lifting, enter AOP-1D, CVCS Malfunction.

B. 1RH-861C, RHR Suction Header Relief valve lifting, enter SEP-2, Shutdown LOCA Analysis.

C. Leak in the RHR heat exchanger aligned for RHR cooling, enter SEP 2, Shutdown LOCA Analysis.

D. Leak in the RCP Seal Return heat exchanger, enter AOP-9B, Component Cooling System Malfunction.

ANSWER:

B.

REFERENCE:

110E018 Sh. 1, P&ID Auxiliary Coolant System, Rev 70 SEP-2, Shutdown LOCA Analysis, Rev 6 NEW HIGHER K/A: 005A1.05: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: Detection of and response to presence of water in RHR emergency sump.

LEARNING OBJECTIVE:

RECOGNIZE the plant conditions under which the SEP-2 series of procedures applies. (031.03.LP2436.001)

RECOGNIZE the entry conditions for and LIST the major actions for the following procedures:

SEP-2 (031.03.LP2436.002)

EXPLANATION:

A. INCORRECT: Relief goes to PRT and would cause lowering RCS pressure but not cause sump A level to rise. This is the correct procedure to enter for the malfunction in this distractor. Plausible if the student has a misconception of the relief discharge path.

B. CORRECT: Relief goes to floor of containment causing sump A level to rise, and RCS pressure to lower. SEP-2 is the required procedure to enter for diagnosing plant conditions in order to transition to another

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 64 of 199 SEP-2 series procedure for mitigating the event.

C. INCORRECT: Leak in this area would cause RCS pressure to lower but a sump in the PAB would rise instead of sump A in containment.

SEP-2 is the required procedure to enter for diagnosing plant conditions in order to transition to another SEP-2 series procedure for mitigating the event. Plausible if the student has a misconception of the leak path.

D. INCORRECT: Leak in this heat exchanger would cause RCS pressure to lower due to being solid but VCT level would rise. This is the correct procedure to enter for the malfunction in this distractor. Plausible if the student has a misconception of the leak path.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 65 of 199 QUESTION # 031 Given the following:

• Unit 1 is in Mode 3 preparing to enter Mode 2

• SI Accumulator parameters are as follows:

o A SI Accumulator Boron concentration - 3050 ppm Level - 43%

Pressure - 810 psig o B SI Accumulator Boron concentration - 2690 ppm Level - 9%

Pressure - 730 psig Which of the following describes the impact, if any, on LCO 3.5.1 Accumulators?

A. Only A SI Accumulator is Inoperable and LCO 3.5.1 is NOT MET.

B. Only B SI Accumulator is Inoperable and LCO 3.5.1 is NOT MET.

C. Both A and B SI Accumulators are Inoperable and LCO 3.5.1 is NOT MET.

D. Both A and B SI Accumulators are OPERABLE and LCO 3.5.1 is MET.

ANSWER:

C.

REFERENCE:

TS 3.5.1, Emergency Core Cooling Systems (ECCS), Accumulators, Rev Amend 241/245 TS B 3.5.1 Bases, Emergency Core Cooling Systems (ECCS),

Accumulators, Rev Amend 241/245 NEW FUNDAMENTAL K/A: 006A1.02: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ECCS controls including: Boron concentration in accumulator, boron storage tanks.

LEARNING OBJECTIVE:

IDENTIFY and DISCUSS Technical Specifications associated with Emergency Core Cooling System components, parameters, and operation including Limiting Condition for Operation (LCO), LCO applicability, Action Conditions and required actions as they pertain to the following requirements:

Accumulators (057.02.LP3340.001)

EXPLANATION:

A. INCORRECT: True, A SI Accumulator is inoperable due to pressure

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 66 of 199 exceeding technical specification values. Plausible if the student cannot correctly recall the required level or boron concentrations.

B. INCORRECT: True, B SI Accumulator is inoperable due to boron concentration lower than technical specification values. Plausible if the student cannot correctly recall the required level or pressure.

C. CORRECT: Both A and B SI Accumulators are inoperable as listed above. LCO 3.0.3 is required to be entered immediately.

D. INCORRECT: A SI Accumulator is inoperable due to pressure exceeding technical specification values and, B SI Accumulator is inoperable due to boron concentration lower than technical specification values. Plausible if the student cannot correctly recall the required pressure or boron concentrations.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 67 of 199 QUESTION # 032 Given the following:

• The reactor is operating at Rated Thermal Power

• The PRT PRESS HI TEMP HI LEVEL HI OR LO alarm is LIT

• The pressurizer PORVs and safety valves have been verified shut The following are indications for the PRT:

• PRT pressure: 6.0 psig

• PRT temperature: 115°F

• PRT level: 72%

Which of the following would cause this alarm condition, and what actions are required to mitigate?

A. Seat leakage from RC-596, PRT Drain Valve to the RCDT.

The PRT will need to be refilled to the required level.

B. Back leakage from the Reactor Coolant System into the PRT.

The PRT will need to be drained and filled to reduce the temperature to the required value.

C. RH-861B, RHR Suction Relief Valve seat leakage into the PRT.

The PRT will need to be drained to the required level.

D. Leakage of nitrogen past valve RC-595, Nitrogen Regulator isolation for the PRT.

The PRT will need to be vented to the required pressure.

ANSWER:

D.

REFERENCE:

STPT 10.1, Setpoint Document, Reactor Coolant System Alarms and Operational Adjustments, Rev 13 541F091 Sh 2, P&ID Reactor Coolant System, Rev 39 ARB 1C04 1C 1-1, 1T-2 PRT Press High Temp Hi Level hi or Low, Rev 5 OP 4C, Pressurizer Relief Tank Operations, Rev 2 MODIFIED: 1999 PBNP NRC Exam FUNDAMENTAL K/A: 007A2.05 Pressurizer Relief Tank / Quench Tank System: Ability to (a) predict the impacts of the following malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: : Exceeding PRT high-pressure limits.

LEARNING OBJECTIVE:

DRAW and DISCUSS a one-line diagram of the Reactor Coolant Drain Tank System similar to TP-2. Discussion of this drawing should include system flowpaths, major components, physical locations, and interfaces with other major systems:

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 68 of 199 RCDT, RCDT Pumps, and Relief valves Containment Isolation valves Inputs to RCDT/System Interfaces

1. Accumulator and Refueling Canal Drains

2. Reactor Vessel and various valve leakoffs

3. RCP Secondary Seal Leakage

4. Excess Letdown Heat Exchanger

5. Pressurizer Relief Tank

6. Containment Sump (051.04.LP0156.002)

EXPLANATION:

A. INCORRECT: Based on the conditions in the PRT, temperature and level both meet the normal required (or non-alarm) conditions, pressure is high. Plausible if the student has the misconception the alarm is due to PRT level, seat leakage from RC-596 would would cause this condition.

B. INCORRECT: Back leakage would cause a rise in temperature not a loss of level. Plausible if the student thinks the alarm is due to a high temperature.

C. INCORRECT: This would cause level to rise, not lower. Plausible is the student thinks the alarm is caused by a high level.

D. CORRECT: Given the condition in the PRT, this is the correct cause of the alarm. The condition would be caused by leakage past RC-595, and would require venting to correct.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 69 of 199 Original Question:

The reactor was operating normally at 100% power when a "PRT PRESS HI TEMP HI LEVEL HI OR LO" alarm activated on the control board. The pressurizer PORVs and safety valves have been verified shut. The following are indications for the PRT:

• PRT pressure: 4.0 PSIG

• PRT temperature: 100°F

• PRT level: 69%

Which of the following would cause this alarm condition?

A. Seat leakage from PRT drain valve RC-596 to the RCDT.

B. Backleakage from the reactor coolant system into the PRT.

C. RHR suction relief valve (RH-861 B) seat leakage into the PRT.

D. Leakage of nitrogen past valve RC-595 into the PRT (assuming the nitrogen pressure regulator valve has been fully opened).

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 70 of 199 QUESTION # 033 Given the following:

• Unit 2 is in Mode 3

• 2P-11A, CCW Pump, is in AUTO after START (running)

• 2P-11B, CCW Pump, is in AUTO after STOP (standby)

• A malfunction causes a trip of 2A05 4160 Safeguards Bus feeder breaker 2P-11A breaker will (1) and 2P-11B will (2) .

(1) (2)

A. trip and need to be reset start on system low pressure when bus voltage returns B. trip and need to be reset get a start signal from the when bus voltage returns UV relays C. remain closed and the start on low system pressure pump will restart when and one pump will need to be voltage is restored secured D. remain closed and the will remain stopped since pump will restart when system pressure will be voltage is restored quickly restored ANSWER:

C.

REFERENCE:

LP0084 Lesson Plan, Component Cooling Water System, Rev 18 883D195 Sh 9, Safeguards Sequence Logic, Rev 19 STPT 8.1, Auxiliary Coolant System Setpoints: General Instrumentation, Rev 8 AOP-18A, Train 'A' Equipment Operation, Attachment A, Rev 15 NEW HIGHER K/A: 008A2.01: Component Cooling Water System: Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of CCW pump.

LEARNING OBJECTIVE:

DESCRIBE the interlocks associated with the Component Cooling Water System and its major components:

Vent valve to atmosphere (CC-17)

CC-769 RCP Pump CCW valves RADWASTE CCW supply and return valves Standby CCW Pump start (051.06.LP0084.004)

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 71 of 199 EXPLANATION:

A. INCORRECT: The standby pump will start on low pressure, but the running pump will actually ride the bus on the loss of power, so it will start up when voltage returns. Plausible because this is how other pumps function in the plant.

B. INCORRECT: The running pump will not trip, it will ride the bus, and the standby pump will start due to low pressure not UV relays.

Plausible because this is how other pumps in the plant function.

C. CORRECT: The running pump will ride the bus, and the standby pump will receive a start signal due to the system seeing a low pressure condition. One pump will need to be secured to restore nomal alignment.

D. INCORRECT: The running pump will ride the bus but the standby pump will receive a start signal due to the system seeing a low pressure condition. Plausible because other pumps in the plant function this way.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 72 of 199 QUESTION # 034 Given the following:

• HC-431K, Pressurizer Master Pressure Controller will NOT control in AUTO.

• HC-431K was placed in MANUAL and demand lowered to 0%.

Pressurizer pressure will...

(Assume no operator actions)

A. RISE until RC-430, PORV OPENS. RC-431C, PORV will remain CLOSED.

B. RISE until RC-431C, PORV OPENS. RC-430, PORV will remain CLOSED.

C. LOWER due to all Control and Backup Heaters remaining OFF.

The Spray Valves remain CLOSED.

D. LOWER due to all Control and Backup Heaters remaining OFF.

The Spray Valves remain OPEN.

ANSWER:

A.

REFERENCE:

LP0078, Power Point, Pressurizer Pressure and Level Control Rev 13 883D195 Sh. 18, Logic Diagram Pressurizer Pressure and Level Control, Rev 13 883D195 Sh. 19, Logic Diagram Pressurizer Heater Control, Rev 5 BANK: 2011 Comanche Peak Exam HIGHER K/A: 010A3.02: Ability to monitor automatic operation of the PZR PCS including: PZR pressure.

LEARNING OBJECTIVE:

DESCRIBE the interlocks, automatic actuation setpoints, and permissives associated with the Pressurizer Pressure and Level Control System. (053.01.LP0457.003)

EXPLANATION:

A. CORRECT: Controller response when placed in MANUAL with a 0%

demand. All Pressurizer heaters turn on and spray valves remain closed. Pressure will rise until 2335 psig when PORV RC-430 will open to lower pressure and RC-431A will remain closed due to the manual signal out of HCV-431K.

B. INCORRECT: This is the opposite of what will happen to the PORVs.

Plausible if the student has a misconception of the effect on the PORVs.

C. INCORRECT: Heaters will turn on. Plausible if the student has a misconception that the heaters will turn off due to the 0% signal placed on HCV-431K.

D. INCORRECT: Heaters will turn on, and spray valve will remain closed.

Plausible if the student has a misconception on the effects of the 0%

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 73 of 199 signal placed on HCV-431K.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 74 of 199 QUESTION # 035 Given the following:

PT-429, Pressurizer Pressure (Red) instrument has been removed from service per 0-SOP-IC-001 Red, Routine Maintenance Procedure Removal of Safeguards or Protection Sensor from Service - Red Channels With the reactor at (1) power, if (2) a reactor trip will be caused?

(1) (2)

A. 30% N-36, IR NI (White) failed HIGH B. 30% TE-403B, THOT (Blue) failed HIGH C. 100% PT-485, Turbine First Stage Pressure (White) failed HIGH D. 100% TE-401B, Loop A Cold Leg Temperature (Red) fails HIGH ANSWER:

B.

REFERENCE:

0-SOP-IC-001 Red, Routine Maintenance Proceudre Removal of Safegurads or Protecting Sensor from Serice - Red Channel, Rev 10 0-SOP-IC-001 Blue, Routine Maintenance Proceudre Removal of Safegurads or Protecting Sensor from Serice - Blue Channel, Rev 12 883D195 Sh 2, Logical Diagrams Treator Trip Signals, Rev 9 883D195 Sh 15, RC Trip signal Logic Diagram, Rev 9 NEW HIGHER K/A: 012A3.06 Ability to monitor automatic operation of the RPS, including: Trip logic.

LEARNING OBJECTIVE:

DIAGNOSE and ANALYZE the effects of and response to Reactor Protection System or associated Process Instrumentation malfunctions/failures. (053.01.LP0315.008)

EXPLANATION:

Removing the pressurizer pressure instrument PT-429 from service will cause the following bistables to be activated (met): Over Temp Trip (OTdeltaT), (Over Temp) Rod Stop (OP/OTdeltaT Turb Runback), High Press Trip (PRZR Hi Press), Low Press Trip (PRZR Lo Press), SI (PRZR Lo Press), and UNBLOCK SI A. INCORRECT: This will not meet the trip logic necessary to cause a reactor trip (power is greater than 25%, this input will be blocked).

Plausible if the student does not recall the power level of 25% to meet

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 75 of 199 the 1 of 2 logic for a reactor trip.

B. CORRECT: This will cause the OTdeltaT Loop B bistable to be activated (met) which meets the trip logic required (2 of 4)

C. INCORRECT: This will not meet the trip logic necessary to cause a reactor trip. Plausible if the student has the misconception that the over power bistable is activated (met) when taking the pressurizer pressure instrument out of service vice the over temperature trip.

D. INCORRECT: This will not meet the trip logic necessary to cause a reactor trip. Plausible if the student has the misconception that this would cause delta to rise, vice lower, meeting the 2 of 4 logic instead of the Thot failure meeting it.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 76 of 199 QUESTION # 036 Given the following:

• Unit 2 is at Rated Thermal Power

• Power Range Channel N41 fails LOW How are power range N41, LOW POW RANGE TRIP and HI POW RANGE TRIP bistables placed in the trip condition?

A. An operator trips the bistables from the instrument racks.

B. The control power fuses are removed from the Power Range N41 drawer.

C. The instrument power fuses are removed from the Power Range N41 drawer.

D. The Comparator Channel Defeat switch is taken to the N41 position at the Comparator and Rate Drawer.

ANSWER:

C.

REFERENCE:

0-SOP-IC-001-RED, Routine Maintenance Procedure Removal of Safeguards or Protection Sensors from Service - Red Channels, Rev 10 MODIFIED: Millstone Unit 3 NRC Exam FUNDAMENTAL K /A: 012A4.05: Reactor Protection System (RPS): Ability to manually operate and/or monitor in the control room: Channel defeat controls.

LEARNING OBJECTIVE:

Describe the response of the Rod Control System to failures and/or malfunctions of the following:

a. Nuclear Instrument (NIS) Power Range Channel

b. Reactor Coolant bypass LOOP RTD failure

c. Turbine First Stage Pressure Transmitter

d. Urgent Failure Alarm EXPLANATION:

A. INCORRECT: This action will not cause the bistables to trip. Plausible because other bistables are tripped using this method.

B. INCORRECT: This action will trip the bistables, but this method is not procedurally allowed. Plausible because the instrument power fuses are utilized to place bistables in the tripped condition.

C. CORRECT: This is the action directed in 0-SOP-IC-001 RED, Attachment A, step 11.

D. INCORRECT: This action will not cause the bistables to trip. Plausible if the student misunderstand the process of defeating a failed channel.

(055.03.LP2441.002)

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 77 of 199 Original Question:

With the plant at 100% power, power range channel N41 fails low.

How are power range N41 HI FLUX and RATE TRIP bistables placed in the trip condition?

A. The control power fuses are removed from the Power Range N41 drawer.

B. The instrument power fuses are removed from the Power Range N41 drawer.

C. The Comparator Channel Defeat switch is taken to the N41 position at the Comparator and Rate Drawer.

D. An I&C technician trips the bistables from the instrument rack room.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 78 of 199 QUESTION # 037 Which of the following describes the expected response to resetting Safety Injection (SI) with an automatic signal STILL present? (Assume all systems functioned as designed, no faults other than the cause of the SI exist, and the SI Reset Pushbuttons are depressed 5 minutes after the SI actuation)

The SI (1) reset, the SI Reset Bypass Activated Lights on CO1R (2) light.

(1) (2)

A. will not will not B. will not will C. will will D. will will not ANSWER:

C.

REFERENCE:

883D195 Sh.7, Logic Diagrams Safeguards Actuation Signals, Rev 25 MODIFIED HIGHER K/A: 013A4.02: Ability to manually operate and/or monitor in the control room: Reset of ESFAS channels.

LEARNING OBJECTIVE:

DESCRIBE/ANALYZE the requirements for the following Safety Injection operations:

c. Resetting automatic safety injection (053.06.LP0486.009)

EXPLANATION:

A. INCORRECT: The SI will reset, and with lights will light. Plausible if the student has a misconception that the SI signal not reset and then the lights and annunciator will remain extinguished.

B. INCORRECT: The SI signal will reset. Plausible if the student has a misconception having to do with the fact the SI signal did not reset, and was bypassed due to the operator action of depressing the reset pushbuttons.

C. CORRECT: The SI will reset and the light and annunciator will illuminate, due to having all necessary power supplies available.

D. INCORRECT: The lights and annunciator will illuminate. Plausible if the student has a misconception of resetting the SI while the signal is still active.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 79 of 199 Original Question:

Which of the following describes the expected response to resetting Safety Injection (SI) with an automatic signal NO longer present?

(Assume SI Reset Pushbuttons are depressed 5 minutes after SI actuation)

A. SI will reset; the SI Reset Bypass Activated Lights on CO1R will light, and C01 A3-6 Unit 1 or 2 SI Train Reset Bypass Activated annunciator will light.

B. SI will reset; the SI Reset Bypass Activated Lights on CO1R will NOT light, and C01 A3-6 Unit 1 or 2 SI Train Reset Bypass Activated annunciator will NOT light.

C. SI will NOT reset; the SI Reset Bypass Activated Lights on CO1R will light, and C01 A3-6 Unit 1 or 2 SI Train Reset Bypass Activated annunciator will light.

D. SI will NOT reset; the SI Reset Bypass Activated Lights on CO1R will NOT light, and C01 A3-6 Unit 1 or 2 SI Train Reset Bypass Activated will NOT light.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 80 of 199 QUESTION # 038 Given the following:

• Both Units were at Rated Thermal Power

• Unit 2 Safety Injection occurs What is the expected response of the Service Water System to the Safety Injection?

A. SW-2930A/2930B, SFP Heat Exchanger Outlet valves close AND SW-2927A/2927B, SFP Heat Exchanger Inlet valves remain open.

B. 2SW-2880, Unit 2 Turbine Hall Service Water Supply valve closes to isolate water to Unit 2 Turbine Hall.

C. SW-2869/2870, Service Water Cross-Connect valves close to isolate the West Service Water Header.

D. 2SW-2907/2908, 2HX-15A-D Containment Recirc Heat Exchanger Emergency flow control valves open to raise flow to Containment Accident Fan Coolers.

ANSWER:

D.

REFERENCE:

883D195 Sh. 8, Logic Diagram Safeguards Sequence, Rev 19 883D195 Sh. 9, Safeguards Sequence Logic, Rev 19 BANK: 2009 PBNP NRC Exam FUNDAMENTAL K/A: 022A4.04: Ability to manually operate and/or monitor in the control room: Valves in the CCS.

LEARNING OBJECTIVE:

List the fans, pumps, and valves, in sequence, which are actuated by timer following an SI actuation. (053.01.LP0486.012)

EXPLANATION:

A. INCORRECT: All SFP heat exchanger cooling MOV's get a shut signal on an SI. Plausible if the student has a misconception and determines that only one isolations shutting is required.

B. INCORRECT: All non-essential SW loads are isolated with an SI on either unit except for the turbine hall cooling. Plausible if the student does not understand that some loads even though they are non-essential, will continue to be supplied.

C. INCORRECT: The main SW header is not isolated in an SI. Plausible because the main SW header has the least amount of loads and redundant safety related loads (accident fan cooling) so it would be the logical choice to isolate if increased SW flow was needed.

D. CORRECT: These are the interlocks which will activate.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 81 of 199 QUESTION # 039 Given the following:

• Unit 1 is conducting a plant start-up per OP 1C, Startup to Power Operation

• The Main Turbine Generator is ready to synchronize to the grid

• A breaker malfunction causes 1W-1B, Containment Accident Fan to be de-energized What impact, if any, will this have on Technical Specifications?

A. None, only 3 of the 4 Containment Accident Fans are required to be operable for the given MODE.

B. LCO 3.6.6 Containment Spray and Cooling Systems is NOT met, take actions to restore 1W-1B back to service.

C. LCO 3.6.6 Containment Spray and Cooling Systems is NOT met, you have one hour to raise cooling water to the other three fans.

D. LCO 3.6.6 Containment Spray and Cooling Systems is NOT met, you have one hour to start all available Accident Fans.

ANSWER:

B.

REFERENCE:

TS 3.6.6 Containment Spray and Cooling Systems, Rev Amend 201/206 TS B 3.6.6 Containment Spray and Cooling System Bases, Rev Amend 241/245 NEW FUNDAMENTAL K/A: 022G2.2.22 Containment Cooling System (CCS): Knowledge of limiting conditions for operations and safety limits.

Cognitive Level:

LEARNING OBJECTIVE:

IDENTIFY and DISCUSS the Technical Specifications associated with the following Containment components, parameters, and operation including Limiting Conditions for Operation (LCO, LCO Applicability, Action Conditions and Required Actions as they pertain to the following requirements:

f. Containment Spray and Cooling Systems (057.01.LP3342.001)

EXPLANATION:

A. INCORRECT: The LCO is not being met. Plausible because other LCOs have similarly worded requirements, TS 3.7.7 for CCW can have one of the 4 heat exchangers OOS and still have the LCO met in Modes 1-4. Normally 3 fans are running, but in the current mode all four fans are required to be operable.

B. CORRECT: The LCO is not met, and these are the required actions.

C. INCORRECT: Correct as far as the LCO not being met, incorrect actions to take. Plausible with the one hour due to many other 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> requirements in Tech Specs, and the automatic increase of cooling

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 82 of 199 water that occurs on an SI signal.

D. INCORRECT: Correct as far as the LCO not being met, incorrect actions to take. Plausible with the one hour due to many other 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> requirements in Tech Specs.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 83 of 199 QUESTION # 040 Given the following:

• Unit 2 is at Rated Thermal Power

• At 0400, both trains of Containment Spray System are determined to be inoperable Which of the following correctly states the required Technical Specification action?

Within one (1) hour, the operators must. . .

A. Return at least one train to service or initiate action to be in Hot Standby by no later than 1000.

B. Initiate action in order to be in Hot Standby by no later than 1100.

C. Return at least one train to service or initiate action to be in Hot Standby by no later than 1600.

D. Initiate action in order to be in Hot Standby by no later than 1700.

ANSWER:

B.

REFERENCE:

TS 3.6.6, Containment Spray and Cooling Systems, Rev Amend 201/206 TS 3.0.3, Limiting Condition for Operation (LCO) Applicability, Rev Amend 215/220 NEW FUNDAMENTAL K/A: 026G2.2.39: Knowledge of less than or equal to one hour Technical Specification action statements for systems.

LEARNING OBJECTIVE:

IDENTIFY and DISCUSS the Technical Specifications associated with Containment Spray System components, parameters, and operation to include:

Limiting Conditions for Operation (LCO)

LCO Applicability (051.03.LP0064.007)

EXPLANATION:

A. INCORRECT: This is incorrect because the amount of time to be in Hot Shutdown is too little. Plausible because the operator should make attempts to correct the issue, but the hour given in LCO 3.0.3 is to make preparations.

B. CORRECT: According to Technical Specification 3.6.6, two independent Containment Spray Systems shall be OPERABLE with each Spray System capable of taking suction from the RWST and manually transferring suction to the containment sump via the RHR System. This Technical Specification is applicable during Modes 1-4.

If both trains are inoperable, the operator must initiate actions per LCO 3.0.3 for manuevering the plant to be in the required mode by the required time.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 84 of 199 C. INCORRECT: This is incorrect because the amount of time to be in Hot Shutdown is too long. Plausible because the operator should make attempts to correct the issue, but the hour given in LCO 3.0.3 is to make preparations.

D. INCORRECT: This is incorrect because the amount of time to be in Hot Shutdown is too long. Plausible because actions should be taken to get the plant to hot standby, and the hours are similar to the hours found in other TS

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 85 of 199 QUESTION # 041 Given the following:

• Unit 1 is at Rated Thermal power

• The reactor tripped due to a failed reactor trip relay What happens to the steam stored in the Main Steam Reheaters (MSRs) and Turbine, and why?

A. Condenser Steam and Crossover Steam Dumps open for about 60 seconds and both continue to operate based on RCS temperature to stabilize the primary plant.

B. The Turbine Generator stays tied to the grid for about 60 seconds and the Crossover Steam Dumps will open to protect the turbine from an overspeed condition if required.

C. The Condenser Steam Dumps open to remove the initial decay heat build up from the primary, steam stored in the MSRs and Turbine remains due to system controls and design.

D. The Turbine Generator is removed from the grid and the Low Pressure Turbine rupture discs remove excess steam energy, protecting the turbine from an overspeed condition.

ANSWER:

B.

REFERENCE:

883D195 Sh 3, Turbine Trip Signals Logic Diagram, Rev 35 FASR, Section 10.1, page 10.1-6 of 50, Rev 2014 NEW FUNDAMENTAL K/A: 039K1.05l: Knowledge of the physical connections and/or cause-effect relationships between the MRSS and the following systems: T/G.

LEARNING OBJECTIVE:

STATE actuation setpoints and EXPLAIN effects of automatic actuations and interlocks associated with the Crossover Steam Dump System (052.02.LP0051.004)

EXPLANATION:

The crossover steam dump system is to provide a means of energy removal from the turbine in the event of a unit trip and it designed to assure that the maximum overspeed of 132% will not be exceeded. The system is armed at 540 MW equivalent load and actuated upon turbine trip at 104% of design speed.

A. INCORRECT: The crossover steam dumps are not designed to maintain the temperature of the primary. Plausible if the student combines the purposes of the condenser and crossover steam dumps.

B. CORRECT: See above.

C. INCORRECT: This is the purpose of the condenser steam dumps, but the steam energy in the MSRs and turbine is dealt with utilizing the crossover steam dumps. Plausible if the student has a misconception

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 86 of 199 of the plant status or automatic plant actions.

D. INCORRECT: The purpose of the rupture disc is to protect the turbine from pressure, not overspeed. Plausible if the student has a misconception on the purpose of the blow out discs.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 87 of 199 QUESTION # 042 Given the following:

• At 10:00, Unit 1 plant conditions were as follows:

o Unit 1 is in MODE 3, cooling down per OP 3C, Hot Standby to Cold Shutdown o RCS TAVG at 495°F and Stable o RCS Pressure is at 1750 psig and Stable o Both Steam Generator pressures are 640 psig and Stable

• At 10:20, Unit 1 plant conditions were as follows:

o A Steam Line Break occurs on A SG Main Steam line inside containment resulting in the following indications:

Containment Pressure is 10 psig and Rising RCS Pressure is 1150 psig and Lowering A SG Pressure is 400 psig and Lowering B SG Pressure is 460 psig and Lowering Both A SG HIGH STEAM FLOW bistables are LIT Both B SG HIGH STEAM FLOW bistables are NOT LIT Which of the following describes the expected position of 1MS-2018, A Main Steam Line Isolation Valve and why? (Assume no operator action)

A. OPEN, the valve will automatically close when Containment Pressure rises to 15 psig.

B. OPEN, the valve will not automatically close due to the SI BLOCK, it must be manually closed.

C. CLOSED, the valve will have automatically closed when 'A' SG Pressure reached 545 psig.

D. CLOSED, the valve will have automatically closed due to the A SG Steam Flow HIGH, and LOW Tavg with a SI signal.

ANSWER:

D.

REFERENCE:

883D195 Sh 7, Logic Diagrams Safeguards Actuation Signals, Rev 25 STPT 2.2, Steam Line isolation, Rev 6 BANK HIGHER K/A: 039G2.4.46: Main and Reheat Steam (MRSS): Ability to verify that the alarms are consistent with the plant conditions.

LEARNING OBJECTIVE:

IDENTIFY and DESCRIBE the Control Room controls, alarms, and indications associated with the Main Steam System, including:

Location and function of component and/or system operating controls

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 88 of 199 and control stations Alarming locations and response to major system and component alarms Plant, system, and component conditions or permissives required for Control Room operation Setpoints associated with major system alarms and/or interlocks (052.02.LP0153.009)

EXPLANATION:

A. INCORRECT: This is incorrect given the conditions. Plausible because the position is correct if the student does not recall the automatic closure interlock of High-High Steam Flow, a Safety Injection Signal, and Low TAVE, and does recall the interlock for containment pressure and the fact that interlock cannot be blocked by the SI Block function.

B. INCORRECT: This is incorrect given the conditions. Plausible because the position would be correct if the student does not recall the automatic closure interlock of High-High Steam Flow, a Safety Injection Signal, and Low TAVE.

C. INCORRECT: This is incorrect given the conditions. Plausible if the student has a misconception of the SI Block function.

D. CORRECT: The conditions show, the automatic closure interlock of High-High Steam Flow, a Safety Injection Signal, and Low TAVE is met and the valve should be closed.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 89 of 199 QUESTION # 043 Given the following:

• Unit 1 is at 20% reactor power

• 1P-28A, Steam Generator Feed Pump, is running

• 1P-28B, Steam Generator Feed Pump, is in standby

• The Main Turbine Generator tripped due to an EHC malfunction What impact, if any, will this have on feeding the steam generators?

1P-28A 1P-28B 1P-53, MDAFW pump A. running standby standby B. tripped running standby C. tripped standby running D. running standby running ANSWER:

A.

REFERENCE:

MDB 3.2.1 1A01, Master Data Book 4160 V AC Unit 1, Rev 7 MDB 3.2.1 1A02, Master Data Book 4160 V AC Unit 1, Rev 6 MDB 3.2.1 1A06, Master Data Book 4160 V AC Unit 1, Rev 2 NEW HIGHER K/A: 059K1.03: Knowledge of the physical connections and/or cause-effect relationship between the MFW and the following systems: S/Gs.

LEARNING OBJECTIVE:

DESCRIBE the interlocks, automatic actuations, and permissive associated with major components of the Feedwater System.

(052.06.LP0128.004)

INDENTIFY and RESPOND to the following failures/transients Feedwater Control System malfunctions Steam Generator Level Control System malfunctions Sudden Load Decrease Steam Generator level increase following feedwater isolation Sudden closure of a feed regulating valve at thigh power (052.02.LP0131.006)

EXPLANATION:

The unit will have a turbine trip with no reactor trip at the given power level. SGFP's have no auto start feature associated with them. 1P-53 MDAFW pump would not get a start signal for low S/G level or AMSAC due to the power level.

A. CORRECT: See above.

B. INCORRECT: See above. Plausible if the student has an incorrect concept of the SGFP having an auto start feature.

C. INCORRECT: See above. Plausible if the student has an incorrect

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 90 of 199 concept of the amount of shrink the steam generators will have with a trip at this power level.

D. INCORRECT: See above. Plausible if the student has an incorrect concept of the amount of shrink the steam generators will have with a trip at this power level.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 91 of 199 QUESTION # 044 Given the following:

• Unit 2 was operating at 100% reactor power

• All systems in their normal at-power alignment

• A failed air start motor caused G-02, Emergency Diesel Generator to be declared out of service

• The alternate emergency diesel power source was aligned to the affected bus Which of the following additional combinations of equipment malfunctions would result in a loss of normal and emergency power availability for the 2P-53, Motor-Driven AFW Pump?

A. 2X-02 lockout and G-01 EDG failure B. 2X-02 lockout and G-04 EDG failure C. 2X-04 lockout and G-01 EDG failure D. 2X-04 lockout and G-04 EDG failure ANSWER:

C.

REFERENCE:

MDB 3.2.2 2A05, Master Data Book 4160 V AC SWGR Unit 2, Rev 11 NEW HIGHER K/A: 061K2.02: Knowledge of bus power supplies to the following: AFW electric drive pumps.

LEARNING OBJECTIVE:

STATE the power supply for the following Auxiliary Feedwater System Components.

Electric-Driven Auxiliary Feedwater Pumps Motor Operated Isolation Valves (052.05.LP0169.003)

EXPLANATION:

A. INCORRECT: 2X-02 supplies the non-safeguard, and 2X-04 supplies the safeguards buses. Plausible if the student has a misconception of how the safeguards/non-safeguard buses are supplied.

B. INCORRECT: The alternate power supple for G-02 is G-01. 2X-02 supplies the non-safeguards, and 2X-04 supplies the safeguards buses. Plausible if the student has a misconception of which EDG is the alternate one based on unit designation and diesel controls placement.

C. CORRECT: The alternate power supple for G-02 is G-01. 2X-02 supplies the non-safeguards, and 2X-04 supplies the safeguards buses D. INCORRECT: The alternate power supple for G-02 is G-01. Plausible if the student has a misconception of which EDG is the alternate one based on unit designation and diesel controls placement

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 92 of 199 QUESTION # 045 Given the following:

• Unit 1 is at Rated Thermal Power

• The Turbine Hall AO reports that the P-38A, Standby Steam Generator Feed Pump, casing is too hot to touch

• 1TR-2000B point 25, P-38A INBD PUMP BRG, is reading 180°F and slowly rising

• The Crew has entered AOP-2C, Auxiliary Feed Pump Steam Binding or Overheating Which of the following is the required action(s)?

A. Manually re-seat P-38A discharge MOVs and monitor temperatures.

B. Check shut P-38A discharge MOVs and start the pump to cool the casing.

C. Start P-38A and vent the casing until a solid stream of water is observed to allow cooling.

D. Check shut P-38A discharge MOVs and vent the pump casing until a solid stream of water is observed to allow cooling.

ANSWER:

D.

REFERENCE:

AOP-2C, Auxiliary Feed Pump Steam Binding or Overheating Unit 1, Rev 10 MODIFIED: 2010 Prairie Island NRC Exam FUNDAMENTAL K/A: 061K5.05: Knowledge of the operational implications of the following concepts as the apply to the AFW: Feed line voiding and water hammer.

LEARNING OBJECTIVE:

Given access to appropriate equipment/indication, DIAGNOSE malfunctions associated with the following systems:

d. Auxiliary Feedwater (055.03.LP2439.001)

EXPLANATION:

A. INCORRECT: This condition is caused by leaking valves. Plausible as one might think re-seating a valve would mitigate the condition.

B. INCORRECT: The AOP has you shut the discharge MOV's. Plausible to think starting the pump with CST water would cool down the pump casing.

C. INCORRECT: Venting the casing is a correct response but the pump is not started. Plausible because this could cause more water to flow allowing a faster cooldown of the pump.

D. CORRECT: AOP has you shut the discharge MOV's and vent the pump casing if <200°F to allow cooling.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 93 of 199 Original Question:

Given the following conditions:

• Unit 1 is at 100% power.

• ERCS point 1T2871A, 12 MD AFWP DISC Temp is reading 180°F and slowly rising.

• Turbine Building operator reports that the 12 MD AFW pump casing is too hot to touch.

• Shift Supervisor enters C28.1 AOP 1, STEAM BINDING OF AN AUXILAIRY FEEDWATER PUMP Which of the following is the required action(s)?

A. Close the affected pump discharge MOV's and start the pump and allow the pump to cool.

B. Start the affected pump and allow the pump to cool, then vent the pump casing until a solid stream of water is observed.

C. Close the discharge MOV's and allow pump to cool, then vent the pump casing until a solid stream of water is observed.

D. Manually re-seat the affected pumps discharge MOV's and allow the pump to cool while monitoring pump casing and discharge temperatures.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 94 of 199 QUESTION # 046 Given the following:

• Units 1 and 2 are at Rated Thermal Power

• An electrical perturbation results in the following:

• 2A52-96, 2A04 to 2A-06 Bus Tie Breaker trips open

• 2B52-25B, 2B04 480 VAC Safeguards Bus normal feeder breaker trips open

• No faults are present on any bus Which of the following describes the availability of electrical power to 2P-15B, B SI pump and 2P-10B, B Residual Heat Removal pump 2 minutes later? (Assume no operator actions) 2P-15B 2P-10B A. available available B. available not available C. not available available D. not available not available ANSWER:

B.

REFERENCE:

499B466 Sh 272, Schematic Diagram, 4160V SWGR Bus 1-A06 (2-A06)

Supply Breaker 1A52-77 (2A52-96) Tie to Bus 1A04 (2A04), Rev 11 499B466 Sh 382, Elementary Wiring Diagram Sta. Serv. Tran. Bkr.

2B52.25B, Rev 18 PBE-7033, Simplified Electrical Power Distribution, Rev 5 MDB 3.2.4 Panel 2B04 Master Data Book, 480 V AC Unit 2, Rev 13 MDB 3.2.2 2A06, Master Data Book 4160 V AC Unit 1, Rev 3 MODIFIED HIGHER K/A: 062K2.01: A.C. Electrical Distribution: Knowledge of bus power supplies to the following: Major system loads.

LEARNING OBJECTIVE:

STATE the loads of the AO buses (054.02.LP0007.012)

EXPLANATION:

Both breakers tripped open with no associated faults on the buses, when 2A06 loses power this will cause G03 and G04 to receive a start signal.

When the EDG reaches rated voltage and frequency it will auto load onto the associated bus. When 2B04 loses power, in order to restore power to the bus the breaker will have to be manually operated. The SI pump is a 4160v load and will have power when the EDG loads onto the associated bus. The RHR pump is a 480v load, and will not have power unless manual actions for either powering the bus back up or supplying

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 95 of 199 the pump from the alternate power supply, as 2B-04 is the normal power supply.

A. INCORRECT: Plausible because the RHR pump has an alternate power supply which must be manually operated. If the student has a misconception concerning the power supply being alternate vice normal.

B. CORRECT: See above.

C. INCORRECT: Plausible because the student may have a misconception about the starting of the EDG given a breaker trip with no fault and the power supply for the RHR as describing in "A" above.

D. INCORRECT: Plausible because the student may have a misconception about the starting of the EDG given a breaker trip with no fault

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 96 of 199 Original Question With both Units at rated power, an electrical perturbation occurs which results in the following:

• 2A52-92, 2A06 4160 VAC Safeguards Bus normal feeder breaker trips open

• 2B52-25B, 2B04 480 VAC Safeguards Bus normal feeder breaker trips open

• No faults are present on any bus Which of the following describes the availability of electrical power to 2P-15B, B SI pump and 2P-14B, B Spray pump 2 minutes later? (Assume no operator actions.)

2P-15B 2P-14B A. available available B. available not available C. not available available D. not available not available

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 97 of 199 QUESTION # 047 According to AOP 0.0, Vital DC System Malfunction, a loss of which of the following would cause a DUAL UNIT TRIP?

A. D01 B. D13 C. D18 D. D21 ANSWER:

A.

REFERENCE:

AOP-0.0, vital DC system Malfunction, Attachment E, Rev 34 PBN LP0121, DC Distribution, Rev 18 BANK: 2012 PBNP NRC Exam (Last 2 NRC Exams)

FUNDAMENTAL K/A: 063K2.01 D.C. Electrical Distribution: Knowledge of bus power supplies of the following: Major DC loads.

LEARNING OBJECTIVE:

Given access to the appropriate equipment indication, DIAGNOSE and DESCRIBE the plant and operator(s) response to the following condition:

Loss of a DC Bus (055.03.LP3456.001)

EXPLANATION:

A. CORRECT: Correct as a loss of D01 (Red) or D02 (Blue) will cause a dual unit trip.

B. INCORRECT: Loss of D13 (Blue) will cause Unit 2 to trip. Plausible is the student cant remember which blue or Red loss will cause a dual unit trip.

C. INCORRECT: Loss of D18 (Blue) will cause Unit 2 to trip. Plausible is the student cant remember which blue or Red loss will cause a dual unit trip.

D. INCORRECT: Loss of D21 (Blue) will cause Unit 1 to trip. Plausible is the student cant remember which blue or Red loss will cause a dual unit trip.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 98 of 199 QUESTION # 048 Given the following:

• The unit suffered a loss of all AC power

• G01 EDG is being started locally

• G01 voltage is out of the normal band due to the field not flashing What action, if any, is needed to flash the field on G01 EDG?

A. The field cannot be manually flashed and the EDG must be shutdown.

B. Depress the FFC relay inside the C-34 cabinet for 3 seconds.

C. Take the EDG Mode Selector Switch in the C-64 cabinet to HYD.

D. Depress the VOLTAGE SHUTDOWN reset pushbutton in the C-81 cabinet.

ANSWER:

B.

REFERENCE:

ECA-0.0, Loss of All AC Power, Attachment A, G01 Local Manual Start, Rev 62 NEW FUNDAMENTAL K/A: 063K3.01: Knowledge of the effect that a loss or malfunction of the DC electrical system will have on the following: ED/G.

LEARNING OBJECTIVE:

Given access to the site specific simulator, IMPLEMENT appropriate actions if the EDG does not start satisfactorily. (031.02.LP0462.007)

IDENTIFY and DESCRIBE the local controls, alarms, and indications associated with the Diesel Generator System, including:

Location and function of component and/or system operating controls and control stations Alarming locations and response to major system and component alarms Plant, system, and component conditions or permissives required for local operation Setpoints associated with major system alarms and/or interlocks (054.02.LP0133.007)

EXPLANATION:

A. INCORRECT: This is an action you take in ECA 0.0 for any EDG if you cannot get a correct response after taking local actions. Plausible as this is done after the attempt to locally start.

B. CORRECT: This is the correct action per ECA 0.0 Attachment A G01 Local Manual Start.

C. INCORRECT: This action is taken for frequency issues or after you manually flash the field. Plausible is the student confuses actions for frequency and voltage.

D. INCORRECT: This is an action you take for G03/G04 EDGs.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 99 of 199 QUESTION # 049 Given the following:

• Unit 1 experienced a trip with a loss of all off-site power

• The unit has been stabilized using required procedures

• Due to equipment issues the steam generators are being fed as follows:

o 'A' S/G being fed by 1P-53 Motor Driven AFW Pump o 'B' S/G being fed by P-38B Standby Steam Generator Pump

• Subsequently, G01 EDG output breaker opened and was re-closed to power up the associated safeguards bus What is the status of the feedwater to the Unit 1 S/Gs?

1P-53 P-38B A. Remains running Is stripped and needs to be manually restarted B. Is stripped and will auto Remains running sequence back on C. Is stripped and will auto Is stripped and needs to be sequence back on manually restarted D. Remains running Remains running ANSWER:

D.

REFERENCE:

MDB 3.2.1 1A06, Master Data Book 4160 V AC SWGR Unit 1, Rev 2 MDB 3.2.4 PANEL 2B04, Master Data Book 480 V AC Unit 2, Rev 12 883D195 Sh. 20, Auxiliary Feedwater Pumps Start-up Logic, Rev 19 883D195 Sh. 25, Auxiliary Feedwater Pumps Start-up Logic, Rev 1 NEW HIGHER K/A: 064K3.02: Knowledge of the effect that a loss or malfunction of the ED/G will have on the following: ESFAS controlled or actuated equipment LEARNING OBJECTIVE:

Given access to the site specific simulator, IMPLEMENT appropriate actions if the EDG does not start satisfactorily. (031.02.LP0462.007)

EXPLANATION:

A. INCORRECT: This electrical transient would not affect the power supplies for these pumps. Plausible because this is a typical result with a EDG breaker closure for P-38B, and 1P-53 will remain running.

B. INCORRECT: This answer would be correct if it was the G03 EDG that had this problem. Plausible because this answer is correct given the other EDG.

C. INCORRECT: This would be a combination of the first two distractors.

Plausible if the student has misconception about the power supplies

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 100 of 199 and interlocks.

D. CORRECT: 1P-53 MDAFW Pump is powered off of 1A06 4160 Safeguards Bus and P-38B SSG Pump is powered off of 2B04 480 Safeguards Bus. Loss and recovery of G01 to 4160 Safeguards Bus 1A05 should have no effect on feeding Unit 1 S/Gs.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 101 of 199 QUESTION # 050 Given the following:

• Unit 2 is at Rated Thermal Power

• Excess Letdown is in service

• 2RE-217, Component Cooling Water Liquid Monitor, goes in to HIGH ALARM Which of the following will automatically shut?

A. CCW-LW-63 and CCW-LW-64, Radwaste System CC Supply and Return B. 2CC-769, Excess Letdown Heat Exchanger Shell Side Outlet Valve C. 2CC-761A and 2CC-761B, RCP Thermal Barrier Outlets D. 2CC-17, Surge Tank Vent Isolation ANSWER:

D.

REFERENCE:

RMSASRB CI 2RE-217, Radiation Monitoring System Setpoint &

Response

Book Channel Information Sheets, CC Water Liquid Monitor, Unit 2, Rev 2

NEW FUNDAMENTAL K/A: 073K4.01: Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following: Release termination when radiation exceeds setpoint.

LEARNING OBJECTIVE:

DESCRIBE the interlocks associated with the Component Cooling Water System and its major components:

1. Vent valve to atmosphere (CC-17)

2. CC-769

3. RCP Pump CCW valves

4. RADWASTE CCW supply and return valves

5. Standby CCW Pump start (051.06.LP0084.004)

EXPLANATION:

A. INCORRECT: Plausible because these valves will be automatically closed by a Unit 2 containment isolation activation, and these valves are only located on Unit 2.

B. INCORRECT: Plausible because this valve will be automatically closed by a Unit 2 containment isolation activation and excess letdown is in service an could be a cause of the indications.

C. INCORRECT: Plausible because these valves are automatically closed by a high flow rate from the associated thermal barrier, and is one of the items monitored when 2RE-217 goes into Alert or high alarm.

D. CORRECT: This valve receives a shut signal when the radiation

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 102 of 199 monitor 2RE-217 reaches High Alarm due to an RCS to CCW system leak.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 103 of 199 QUESTION # 051 Given the following:

• Unit 1 has been operating at 100% power for several weeks

• 1SC-938C, 1RE-109 Failed Fuel Monitor Flow Throttle, which is normally in a throttled position, was inadvertently placed in a full open position after Reactor Coolant System sampling 1RE-109, Failed Fuel Monitor, indication will...

A. lower due to increased bypass flow around the detector B. rise because more N 16 gammas will reach the detector C. remain the same because the gross specific activity of the reactor coolant has not changed D. lower because of the increased flow which lowers the time N 16 gammas are in the detector ANSWER:

B.

REFERENCE:

RMSASRB CI 1RE-109, Radiation Monitoring System alarm Setpoint &

Response Book Channel Information Sheets, Failed Fuel Monitor Unit 1, Rev 5 BANK: 2009 PBNP NRC Exam HIGHER K/A: 073K5.01: Knowledge of the operational implications as they apply to concepts as they apply to the PRM system: Radiation Theory, including sources, types, units, and effects.

LEARNING OBJECTIVE:

DESCRIBE the function and/or purpose, design bases, and operating characteristics of the Radiation Monitoring System. Description should include:

1. Major component and their physical location

2. Tasks and control functions performed by a DAM and a SPING

3. Operation of an AMAU and AMBU

4. Type of detector utilized and its operation for each RMS channel (053.05.LP0286.001)

EXPLANATION:

A. INCORRECT: 1-SC-938C is inline, not a bypass flow control valve.

Plausible if the student assumes it controls bypass flow.

B. CORRECT: Since this is causing more flow, not less, more flow would result in more N16 passing near the detector and therefore in a higher reading. (RMSASRB CI 1RC-109, high alarm possible causes)

C. INCORRECT: With the more flow, a rise in detector indication is caused. Plausible is the student does not understand N16, even though the gross specific activity has not changed, since the flowrate has, and with that the N16 seen by the detector.

D. INCORRECT: Given the detector and volumetric flow, more flow, more

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 104 of 199 N16, a higher reading. Plausible if the student does not understand that concept, and visualizes it as a time at detector concept, the slow the flow, the higher the reading of activity.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 105 of 199 QUESTION # 052 Given the following:

• Both Units were at Rated Thermal Power

• P-32A, P-32B, and P-32D, Service Water pumps were running

• Subsequently, 1X04, LV Station Aux Transformer locks out

• Unit 1 Safety Injection occurred simultaneously with the 1X04 lockout

• G03 EDG Output breaker failed to close Which of the following is a COMPLETE list of which Service Water pumps will be running two minutes later? (Assume no operator action)

A. A, B, and F B. D, E and F C. A, B, D and E D. A, B, D, E and F ANSWER:

D.

REFERENCE:

883D195 Sh 8, Logic Diagrams Safeguards Sequence, Rev 19 MODIFIED: 2007 PBNP NRC Exam HIGHER K/A: 076K4.02: Knowledge of SWS design feature(s) and/or interlock(s) which provide for the following: Automatic start features associated with SWS pump controls LEARNING OBJECTIVE:

ASSESS the response of the service water system to a Safeguards actuation. (051.06.LP0086.008)

EXPLANATION:

Power supplies to the SW pumps are as follows:

A - 1B03 B - 1B03 C - 1B04 D - 2B04 E - 2B04 F - 2B03 A. INCORRECT: If an 'A' train Diesel starts and loads onto the bus, A, B and F SW pumps would start. This would be in addition to D pump which would remain running. Plausible because this would leave only

'A' train pumps running.

B. INCORRECT: This is not correct given the conditions. Plausible if the examinee believes that only Unit 2 pumps would remain running.

C. INCORRECT: This is not correct given the conditions. Plausible if the examinee erroneously assigned F pump to Unit 1, another common misconception, this would be correct.

D. CORRECT: Correct, as noted in the table above, only C SW pump would be without power, thus on SI all pumps would start, but only C would have no power.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 106 of 199 Original Question Consider the following plant conditions:

• Both Units were at 100% reactor power with normal electric plant lineup.

• P-32A, P-32B and P-32D Service Water pumps were running.

• Subsequently, 1X04, LV Station Aux Transformer locks out.

• Unit 1 Safety Injection occurred simultaneously with the 1X04 lockout.

• G03 EDG Output breaker failed to close.

Which of the following is a COMPLETE list of which Service Water pumps will be running two minutes later? (Assume NO operator action)

A. A, B, D and F B. A, D, E and F C. A, B, C, D and E D. A, B, D, E and F

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 107 of 199 QUESTION # 053 Given the following:

• Unit 1 is at Rated Thermal Power

• A Loss of Instrument Air has occurred

• South Instrument Air Header is reading 0 psig Which of the following describes the effect of the loss of control air on the Unit 1 Main Steam Isolation Valves (MSIVs)?

The MSIVs. . .

A. will remain open and can only be closed locally by actuating the vent solenoid pushbuttons.

B. will remain open due to instrument air not being required to hold the MSIV open.

C. close due to loss of instrument air pressure holding the MSIV open.

D. close due to the vent solenoids no longer being held closed by instrument air pressure.

ANSWER:

C.

REFERENCE:

M-201 Sh1, P&ID Main & Reheat Steam System, Rev 62 MODIFIED: 2008 DC Cook NRC Exam FUNDAMENTAL K/A: 078K1.05: Knowledge of the physical connections and/or cause-effect relationships between the IAS and the following systems: MSIV air.

LEARNING OBJECTIVE:

RECOGNIZE and ANALYZE the effects of the following conditions:

Loss of Instrument Air on Main Steam system operation Improper Main Steam valve lineups on Auxiliary Feedwater Pump operation Loss of Main Steam on Gland Sealing system (052.02.LP0153.006)

EXPLANATION:

A. INCORRECT: Plausible if the student has a misconception on the way the valve operator operates, this is a method for closing the valve, by not the only way given the initial conditions.

B. INCORRECT: Plausible if the student has a misconception on the way the valve operator operates, and there is a spring to assist in the shutting of this valve, incase of a loss of instrument air.

C. CORRECT: The valve will drift off of the open seat due to the loss of instrument air, and after it starts to drift, the interlocks will immediate cause it to shut, and in the event of a complete loss of instrument air, the valve operator is spring assisted, as well as the fact that steam flow will aid in the closure of the valve.

D. INCORRECT: Plausible because this is the manner in which the

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 108 of 199 operator will act on a loss of power. The solenoid valve will go to the fail position, and that will cause that area of piping to vent and the valve to close.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 109 of 199 Original Question:

Given the following plant conditions on Unit 1:

• The unit was operating at 100% power.

• A Loss of Control Air has occurred.

• Control Air Header is reading 0 psig.

Which ONE of the following describes the effect of the loss of control air on the Main Steam Isolation Valves (MSIVs). The MSIVs:

A. will remain open and can be closed using the hydraulic unit B. will remain open and can be closed by locally aligning nitrogen to one MSIV dump valve.

C. close due to MSIV dump valves failing open on loss of air D. close due to loss of control air to the hydraulic unit.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 110 of 199 QUESTION # 054 Given the following:

• Unit 1 is at 28% reactor power

• K3A, Service Air Compressor is running

• 1A04, 4160 Non-Safeguards Bus lockout has occurred, causing a momentary loss of power to 1A06, 4160 Safeguards Bus

• Required EDG is running and supplying 1A06

• The crew is restoring previously running loads What actions, if any, are required to restore K3A Service Air Compressor to service?

A. Nothing, K3A is unaffected by the loss of 1A04 B. Nothing, K3A restarts upon re-energizing of 1B04 C. Reset the K3A control switch from OFF to CONSTANT, at the C01R D. Close the breaker on 1B04 by pressing the CLOSE button, then start K3A ANSWER:

D.

REFERENCE:

AOP-5B, Loss of Instrument Air, NOTE and Step 3, Rev 41 MDB 3.2.3 PANEL 1B04, 480 V AC Unit 1, Rev 14 BANK HIGHER K/A: 078K4.01: Knowledge of IAS design feature(s) and/or interlock(s) with provide for the following: Manual/automatic transfers of control.

LEARNING OBJECTIVE:

IDENTIFY and DESCRIBE the local controls, alarms, and indications associated with the Instrument and Service Air Systems, including:

Location and function of component and/or system operating controls and control stations Alarming locations and response to major system and component alarms Plant, system, and component conditions or permissive required for local operation Setpoints associated with major system alarms and/or interlocks (052.06.LP0338.008)

EXPLANATION:

A. INCORRECT: K3A will be affected by the loss of 1A04. Plausible if this student has a misconception of the interlock at the breaker, which requires manual local operation when an SI (loss and regain of power) happens.

B. INCORRECT: K3A will not restart after 1B04 until local operator action at the breaker happens. Plausible if the student does not recall the local operator actions required.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 111 of 199 C. INCORRECT: There are only indicating lights in the control room.

Plausible if the student cannot recall the which compressors can be operated from the control room.

D. CORRECT: This is the correct sequence of events per AOP-5B NOTES and Step 3

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 112 of 199 QUESTION # 055 Given the following:

• Unit 1 is in Mode 6

• RCS temperature is 100°F and Stable

• RCS time to boil is 1.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />

• Containment operability is not being maintained

• Containment purge is in operation per OP 9C, Containment Venting and Purging

• The Unit 1 containment upper personnel airlock has malfunctioned such that neither door can be shut

• Maintenance estimates that it will take 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> to return at least one of the doors to service Which of the following is a valid concern about the status of the upper airlock?

A. An unmonitored release to the atmosphere is taking place while both airlock doors are open.

B. The lower airlock cannot be utilized because one of its bulkhead doors must be locked shut.

C. The containment closure time requirements of CL 1E, Containment Closure Checklist, are not met.

D. The upper airlock will not be returned to service before the one hour Technical Specification Action Condition time has expired.

ANSWER:

C.

REFERENCE:

CL 1E, Containment Closure Checklist Unit 1 BANK: 2009 PBNP NRC Exam HIGHER K/A: 103K3.01: Knowledge of the effect that a loss or malfunction of the containment system will have on the following: Loss of containment integrity under shutdown conditions.

LEARNING OBJECTIVE:

DESCRIBE the procedures which govern operation of the Containment Structure. Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed and Non-Licensed Operators.

(051.05.LP0099.004)

EXPLANATION:

A. INCORRECT: Not a concern unless a radiological accident occurs, both doors are routinely kept open during periods of outage. Plausible if the student determines that the loss of the closure capability is an issue.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 113 of 199 B. INCORRECT: This is a possible solution if the plant is > 200°F, and there is a need or requirement to isolate the broken door. Plausible if the student does not recall the >200°F portion.

C. CORRECT: This is true, when containment operability is not being maintained.

D. INCORRECT: This is only correct for plant temperature 200°F, so would not be a concern at this time. Plausible if the student does not recall the >200°F portion.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 114 of 199 QUESTION # 056 Given the following:

• Unit 1 is at Rated Thermal Power

• Train 'B' reactor trip breaker testing in progress with the following line up:

• 'A' Reactor Trip Breaker (RTA) is shut

• 'B' Reactor Trip Bypass breaker (BYB) is shut

• An ACTUAL reactor trip condition occurs

• The 'A' Reactor Trip Breaker (RTA) UV coil does not function Which of the following, if any, would be the FIRST to open the the 'A' Reactor Trip Breaker (RTA):

A. Breaker would not trip B. RTA shunt trip coil actuates C. RTA redundant UV coil actuates D. Local breaker trip pushbutton is pressed ANSWER:

B.

REFERENCE:

883D195 Sh. 2, Logic Diagrams Reactor Trip Signals FASR, Section 7.2 Rector Protection System, Rev 2014 BANK FUNDAMENTAL K/A: 001K6.03: Knowledge of the effect of a loss or malfunction on the following CRDS components: Reactor trip breakers, including controls.

LEARNING OBJECTIVE:

DESCRIBE the function and/or purpose, design bases, and operating characteristics for the Process Instrumentation and signal development associated with the Reactor Protection System. Description should include:

1. Major Process Instruments

2. Reactor Coolant Temperature (DeltaT and Tavg)

3. Pressurizer Pressure and Level

4. Reactor Coolant System Flow

5. Steam Flow

6. Feedwater Flow

7. Steam Pressure

8. Steam Generator Level

9. Signal development for the Major Process Instruments

10. Protection/Control Systems which receive inputs from the Major Process Instruments

11. Channel Testing and Tripping (053.02.LP0315.002)

EXPLANATION:

A. INCORRECT: The breaker would trip. Plausible if the student has a

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 115 of 199 misconception on the functions of the reactor trip breaker.

B. CORRECT: This is what would cause the breaker to open.

C. INCORRECT: This would not trip the breaker. Plausible if the student as a misconception of the UV Coils, there are two, and one is a redundant trip feature for the breaker, but they are train specific.

D. INCORRECT: This would trip the breaker, but would not be the first thing to cause that trip. Plausible because this would cause a trip of the bypass breaker.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 116 of 199 QUESTION # 057 Given the following:

• Unit 2 is at Rated Thermal Power

• Volume Control Tank (VCT) level is at 38%

• 2LT-428, Pressurizer Level transmitter (Blue) loses the level in its reference leg

• No ESF actuations occur as a result of the reference leg leak Which of the following describes the VCT Level response as the plant stabilizes? (Assume no operator actions)

A. rise and be maintained between 56% and 78%.

B. lower, then be maintained between 17% and 28%.

C. lower to 4%, cause charging pump suction to transfer to the RWST.

D. rise to between 56% and 78%, then lower and be maintained between 17% and 28%.

ANSWER:

D.

REFERENCE:

LP0079, Pressurizer Pressure and Level Control, Rev 13 Power Point LP0079, Chemical and Volume Control, Rev 24 STPT 7.1, Setpoint Document Chemical and Volume Control System Setpoint: General Instrumentation, Rev 12 883D195 Sh 18, Logic Diagram Pressurizer Pressure and Level Control, Rev 13 BANK: 2009 Sequoyah NRC Exam HIGHER K/A: 011A1.03: Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the PZR LCS controls including: VCT level.

LEARNING OBJECTIVE:

DESCRIBE the automatic functions and interlocks associated with the Pressurizer, Level Control, Pressure Control and Relief System and its major components:

1. Pressurizer Pressure

2. Pressurizer Level

3. Low Temperature Over Pressure Protection (051.01.LP0078.004)

DESCRIBE the interlock associated with the Chemical Volume and Control (include administrative limitations)

1. Orifice Isolation valves

2. Letdown isolation valves

3. Divert valve

4. Volume Control Tank outlet valve and Refueling water makeup supply valves (CV112B and CV112C

5. Containment Isolation Valves

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 117 of 199

6. Charging pumps (051.02.LP0079.004)

EXPLANATION:

A. INCORRECT: VCT level will rise and start to divert as level reaches 56%, but when the pressurizer level drops, and letdown isolates level should go down. Plausible because the VCT level rising and causing the divert valve to open is correct, but with the letdown isolation, will not be maintained.

B. INCORRECT: VCT level will not drop. Plausible because of the common misconception of the level transmitter and reversing the effect of a loss of a reference leg, which would mean that pressurizer level is indicating low, and charging would increase and VCT level would in fact drop.

C. INCORRECT: VCT level will not drop. Plausible because of the common misconception of the level transmitter and reversing the effect of a loss of a reference leg, which would mean that pressurizer level is indicating low, and charging would increase and VCT level would in fact drop, 4% is plausible because that is the point were automatic switchover will occur.

D. CORRECT: The loss of the reference leg will cause the indicated level to fail high, and this is the controlling channel for pressurizer level. With level failing in this manner, charging will slow trying to reduce level, causing a rise in VCT level, maintaining level at 56% to 78%. The reduction of charging flow will cause pressurizer level to lower until letdown isolates on interlock. When letdown isolates the VCT will lose that source of water, and level will now lower and be maintained at 17% to 28%

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 118 of 199 QUESTION # 058 Given the following:

• The reactor is initially critical below the Point of Adding Heat

• Both Reactor Coolant Pumps are running

• Main Steam Isolation Valves (MSIV) and MSIV bypass valves are closed

• Moderator Temperature Coefficient is negative

• RCS TAVE is 547°F and stable with heat removal through the Atmospheric Steam Dumps

• Intermediate Range Nuclear Instruments N-35 and N-36 indicate 4x10-7 amps and Stable If both Atmospheric Steam Dumps fail CLOSED and NO operator action is taken, how will reactor power respond?

Reactor power will lower...

A. initially, then return to 4x10-7 amps on Intermediate Range instruments due to subcritical multiplication.

B. to a subcritical power level and stabilize at a value in the source range based on subcritical multiplication.

C. initially, then return to 4x10-7 amps on Intermediate Range instruments due to the resulting rise in RCS TAVE.

D. until the positive reactivity from the temperature rise causes the reactor to stabilize at a critical power level in the Source Range.

ANSWER:

B.

REFERENCE:

NUC-GFP-RXT-008, Reactor Operational Physics, Rev 0 BANK: 2011 Kewaunee NRC Exam HIGHER K/A: 015K5.06: Knowledge of the operational implications of the following concepts as they apply to the NIS: Subcritical multiplications and NIS indications.

LEARNING OBJECTIVE:

DESCRIBE reactor response once criticality is obtained.

(NUC.GFP.RXT.008.012)

EXPLANATION:

A. INCORRECT: With the atmospheric dump failed shut, the temperature of the RCS will rise, negative reactivity will be inserted due to the negative temperature coefficient, this will cause power will lower. This process makes the return to the original power level wrong. Plausible if the student has a misconception on the effects of the negative temperature coefficient and its effects on reactor power.

B. CORRECT: As temperature rises, negative reactivity is inserted

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 119 of 199 because of the negative temperature coefficient until power is reduced to the source range where subcritical multiplication will take over.

C. INCORRECT: Power will go down, not up with a rise in RCS temperature based on the negative temperature coefficient. Plausible if the student has a misconception on the effects of the negative temperature coefficient.

D. INCORRECT: Subcritical multiplication will cause power to stabilize in the source range, not the positive reactivity caused by the temperature change. Plausible if the student has a misconception concerning subcritical multiplication.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 120 of 199 QUESTION # 059 Given the following:

• Unit 2 is at Rated Thermal Power

• Containment Forced Vent is in progress per OP 9C, Containment Venting and Purging

• 2P-707B, Containment Forced Vent Pump, is running.

IF 2RE-212, Unit 2 Containment Noble Gas Monitor, goes to the HIGH ALARM condition, which of the following automatic actions will occur AND what actions will the Unit 2 operators take?

(CVI - Containment Ventilation Isolation (CI - Containment Isolation)

A. No automatic actions will occur. Operators will compare 2RE-212 readings to other Containment radiation monitors and determine whether forced vent may continue.

B. CVI will automatically occur. Operators will need to manually secure 2P-707B, to prevent it from running without a discharge path.

C. CVI will automatically occur. Operators will verify that 2P-707B, is off and that 2RM-3200H, Containment Forced Vent Pump Discharge Valve, is closed.

D. CI and CVI will automatically occur. Operators will verify 2P-707B, is off and 2RM-3200H, Containment Forced Vent Pump Discharge Valve is closed and that all Containment Isolation Valves repositioned as required.

ANSWER:

C.

REFERENCE:

STPT 13.4, Setpoint Document, Radiation Monitoring System: Effluent Monitors, Rev 16 RMSASRB CI 2RE-212, Radiation Monitoring System Alarm Setpoint &

Response Book Channel Information Sheet, Containment Noble Gas Monitor, Rev 8 ARB C01 C 2-8, Containment Ventilation Isolation BANK HIGHER K/A: 029A3.01: Ability to monitor automatic operation of the Containment Purge System including: CPS isolation.

LEARNING OBJECTIVE:

DESCRIBE the interlocks associated with the Containment Ventilation System and its major components:

1. Purge Exhaust and Supply Fans

2. Steam Generator Channel Head Exhaust Fan

3. Refueling Surface Exhaust and Supply Fans

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 121 of 199

4. Containment Ventilation Supply and Exhaust Fan Dampers

5. Safety Injection Lock Outs

6. Radiation Monitor Actuations (051.05.LP0057.004)

DESCRIBE the sample flowpath for the following modes of operation associated with the Containment Vent Monitor:

1. Containment Sampling

2. Containment Sample with forced Ventilation

3. Stack Sampling

4. Purge

5. Grab Sampling (053.05.LP0286.011)

EXPLANATION:

A. INCORRECT: This is a correct response for operator actions, if no automatic action happened, but CVI will automatically happen.

Plausible if the student only recalls the actions of the RMSASRB CI 2RE-212, as they do not address the automatic CVI.

B. INCORRECT: The automatic operation is correct, but the operator actions are incorrect, as there is no fault called out in the stem pertaining to 2P-707B. Plausible if the student does not recall that 2P-707B will automatically stop, and additional actions are required.

C. CORRECT: CVI will automatically occur, and the operator verifies 2P-707B is off, and will also have to verify 2RM-3200H is closed.

D. INCORRECT: CI will not automatically happen. The operator actions are correct if CI would have automatically happened. Plausible if the student believes that CI will also occur.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 122 of 199 QUESTION # 060 Which of the following describes when the Spent Fuel Bridge Hoist should be placed in 'bypass'?

A. To allow the bridge to access new fuel elevator.

B. To bypass the maximum weight limit interlock of the hoist when raising the hoist.

C. To allow the bridge to use the offset fuel handling tool in the SFP perimeter racks.

D. To bypass the minimum weight limit interlock of the hoist when lowering the hoist.

ANSWER:

D.

REFERENCE:

OI 91, Spent Fuel Bridge Hoists, Rev 27 BANK FUNDAMENTAL K/A: 034G2.1.28: Fuel Handling Equipment System (FHES):

Knowledge of the purpose and function of major system components and controls.

LEARNING OBJECTIVE:

DESCRIBE the interlocks associated with the Spent Fuel Equipment Spent Fuel Pool Bridge (112.01.LP0260.004)

DESCRIBE the procedures which govern operation of the Spent Fuel Pool equipment. Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by licensed and non-licensed operators.

(112.01.LP0260.005)

EXPLANATION:

A. INCORRECT: The interlock is for bypassing the low load limit.

Plausible as the interlock nomenclature does not have a descriptive title. There is no interlock to prevent this.

B. INCORRECT: The interlock is for bypassing the low load limit.

Plausible as the interlock nomenclature does not have a descriptive title and other bypasses exist to bypass the upper limits weight limits.

C. INCORRECT: The interlock is for bypassing the low load limit.

Plausible as the interlock nomenclature does not have a descriptive title and other bypasses exist to allow movement in restricted areas for special handling requirements.

D. CORRECT: This is the reason referenced in the procedure.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 123 of 199 QUESTION # 061 Given the following:

• Unit 1 is operating at Rated Thermal Power

• 1PT-486, HP First Stage Turbine Pressure Transmitter fails as is If the turbine were to trip, the transmitter failure will...

(Assume no operator action)

A. have no effect, the condenser steam dumps will operate as designed causing RCS TAVE to lower to a value of 547°F B. prevent the condenser steam dump valves from "arming", the deviation signal caused by the rising RCS temperature will have no effect C. cause all of the steam dumps to "blow open"; the operator must place the mode selector switch to MANUAL to prevent an excessive RCS cooldown D. prevent the condenser steam dump valves from modulating; RCS TAVE will be maintained at the lowest "blow open" temperature value of 547°F ANSWER:

B.

REFERENCE:

LP0035 Lesson Plan, Condenser Steam Dumps (I&C) Rev 16 BANK FUNDAMENTAL K/A: 041K1.05 Steam Dump System (SDS) and Turbine Bypass Control:

Knowledge of the Physical connections and/or cause-effect relationships between the SDS and the following systems: RCS.

LEARNING OBJECTIVE:

STATE actuation setpoints and EXPLAIN effects of automatic actuations and Interlocks associated with the Steam Dump System. Explanation should include the three modes of Steam Dump operation, Condenser Availability signal, Arming signals, and operation of the four Dump Valve Solenoid Valves. (052.02.LP0035.004)

EXPLANATION:

A. INCORRECT: See the correct answer. Plausible if the student has a common misconception on the operation of the steam dumps, and how the air is supplied to the dumps.

B. CORRECT: This is correct, the loss of the signal from the turbine impulse chamber pressure will cause the steam dump system to not be supplied with air. Without this air being supplied the steam dump valves will not operate in either modulate or trip open mode.

C. INCORRECT: See the correct answer. Plausible as the steam dumps will receive one half of the blow open or trip open signal with the trip of the turbine and build up of decay heat. The other half of the signal is the air signal from the turbine impulse chamber pressure. Also if the

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 124 of 199 reference temperature we to be effect in a similar manner as the pressure transmitter was effected, then the dumps blowing open and an excessive cooldown would be plausible.

D. INCORRECT: See the correct answer. Plausible if the student has a misconception on the logics of the steam dump. During normal operations the dumps would blow open and then modulate at the 547°F value.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 125 of 199 QUESTION # 062 Given the following:

• The Unit is at Rated Thermal Power

• Main Condenser Vacuum has degraded several inches to a new steady state value

• The Crew is responding to the event per the Abnormal Operating Procedures

• Vacuum was recovered by stopping condenser air in-leakage, this was the only action taken to recover vacuum Which of the following will result as vacuum recovers?

A. Rising megawatt output with lowering hotwell temperature.

B. Rising megawatt output with rising hotwell temperature.

C. Lowering megawatt output with lowering hotwell temperature.

D. Lowering megawatt output with rising hotwell temperature.

ANSWER:

A.

REFERENCE:

NUC-GFP-CMP-003, Heat Exchangers and Condensers, Rev 0 NEW HIGHER K/A: 055K3.01 Condenser Air Removal System (CARS): Knowledge of the effect that a loss or malfunction of the CARS will have on the following: Main condenser.

LEARNING OBJECTIVE:

Given access to the Site Specific Simulator or specific plant conditions, EVALUATE plant indications associated with the following events:

1. Secondary Coolant System leak

2. Feedwater system malfunction

3. Loss of Condenser Vacuum

4. Loss of Instrument Air Trip (055.03.LP2439.004)

EXPLANATION:

A. CORRECT: With the condenser starting at a lower (degraded vacuum), as vacuum improves, this will cause megawatt output to rise due to an improved delta H. The condenser is a saturated system, so as vacuum improves, pressure lowers since PSAT and TSAT are related, a lower saturation pressure means a lower saturation temperature.

B. INCORRECT: The second half is incorrect. Plausible if the student reverses the change in absolute pressure, based on the system being a vacuum instead of a pressure.

C. INCORRECT: The first part is incorrect. Plausible if the student reverses the change in absolute pressure, based on the system being a vacuum instead of a pressure.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 126 of 199 D. INCORRECT: Both parts are incorrect. Plausible if the student reverses this for a loss of vacuum vice a recovery of vacuum.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 127 of 199 QUESTION # 063 Given the following:

• A discharge of the A Monitor Tank is in progress Which of the following would provide the Control Operator indication that WL-18, Waste Condensate Overboard Discharge to SW Header Control Valve, should have automatically closed?

A. The COMMON PROCESS RADIATION MONITOR HIGH annunciator alarms.

B. RE-218, Waste Disposal System Liquid Monitor, status indication on the RMS server changes from green to red.

C. RE-223, Waste Distillate Tank Overboard Monitor, status indication on the RMS server changes from green to blue.

D. The status light for WL-18, Waste Condensate Overboard to SW header control valve is lit on the Containment Isolation Panel.

ANSWER:

B.

REFERENCE:

RMSASRB CI RE-218, Radiation Monitoring System Alarm Setpoint &*

Response Book Channel Information Sheet Waste Disposal System Liquid Monitor, Rev 4 RMSASRB CI RE-223, Radiation Monitoring System Alarm Setpoint &*

Response Book Channel Information Sheet Waste Distillate Tank Overboard Monitor, Rev 4 BANK: 2003 PBNP NRC Exam FUNDAMENTAL K/A: 068A4.04 Liquid Radwaste System (LRS): Ability to manually operate and/or monitor in the control room: Automatic isolation.

LEARNING OBJECTIVE:

IDENTIFY and DESCRIBE the controls, alarms, and indications associated with the Liquid Waste Disposal System, including:

1. Location and function of component and/or system operating controls and control stations

2. Alarming locations and response to major system and component alarms

3. Plant, system, and component conditions or permissives required for operation

4. Setpoints associated with major system alarms and/or interlocks (051.04.LP0063.005)

EXPLANATION:

A. INCORRECT: This is incorrect due to this alarm can be caused by a number of monitors, but not RE-218. Plausible is the operator assumes this alarm will be caused by RE-218 as it is a process monitor, an is the only indication in the control room.

B. CORRECT: This is the indication in the control room.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 128 of 199 C. INCORRECT: RE-223 will cause an automatic operation if a high alarm condition exists, the indication on RMS server indicated that the point has gone bad. Plausible if the student expects a bad input to cause a closure signal, and does not understand the system layout.

D. INCORRECT: There are Liquid Waste system valve indications on the Containment Isolation panel, but WL-18 is not one of them. Plausible if the student assumes the isolation panels will indicate for overboard discharges.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 129 of 199 QUESTION # 064 What automatic actions occur when a RE-214, Auxiliary Building Vent Stack Monitor, goes into high alarm and why?

A. Shifts the Auxiliary Building ventilation to 100 % recirculation to minimize radioactive release from the PAB.

B. All PAB ventilation fans trip and associated dampers close to isolate any potential release to outside the PAB.

C. W-35, PAB Supply Fan stops to create a more negative pressure in the PAB to minimize radioactive release to the outside environment.

D. F-23, PAB Charcoal Filter isolation dampers open and F-29, PAB HEPA Filter isolation damper closes to filter and minimize the release.

ANSWER:

D.

REFERENCE:

RMSASRB CI RE-214, Auxiliary Building Vent Exhaust Gas Monitor, Rev 8 AOP-11C, PAB High Airborne, Rev 8 BG AOP-11C, Background Document PAB High Airborne, Rev 6 FUNDAMENTAL BANK K/A: 072K4.03: Knowledge of the ARM system design feature(s) and/or interlocks(s) which provide for the following: Plant ventilation systems LEARNING OBJECTIVE:

DESCRIBE the interlocks associated with the Auxiliary and South Service Building Ventilation Systems and its major components.

(051.05.LP2711.004)

EXPLANATION:

A. INCORRECT: This is not part of the interlocks for this monitor.

Plausible as this action would minimize the release.

B. INCORRECT: This is not the interlock for this monitor. Plausible as this action would isolate potential released which is a normal system interlock.

C. INCORRECT: This is not the interlock for this monitor. Plausible because this is one of the normal mechanisms to prevent or minimize releases.

D. CORRECT: This is the system interlock as called out by RMSASRB CI RE-214.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 130 of 199 QUESTION # 065 Given the following:

• A Loss of Instrument Air has occurred

• The crew is performing actions of AOP-5B, Loss of Instrument Air

• Instrument Air header pressure is 80 psig and Lowering Slowly Which of the following is the required action related for IA-3079 and IA-3014, North and South Header Service Air to Instrument Air Backup valves?

A. Ensure IA-3079 and IA-3014 automatically opened; the valves opened at 85 psig and lowering.

B. Manually open IA-3079 and IA-3014; the valves have no automatic features.

C. Ensure IA-3079 and IA-3014 automatically closed; the valves closed at 85 psig lowering.

D. Manually close IA-3079 and IA-3014; the valves have no automatic features.

ANSWER:

A.

REFERENCE:

LP0338 Instrument Service and Emergency Breathing Air Systems, Rev 14 M-209 Sh 3, P&ID Instrument Air, Rev 18 AOP-5B, Loss of instrument Air, Rev 41 BANK: 2007 McGuire NRC Exam FUNDAMENTAL K/A: 079A2.01: Ability to (a) predict the impacts of the following malfunctions or operations on the Station Air System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations Cross-connection with IAS.

LEARNING OBJECTIVE:

IDENTIFY and DESCRIBE the local controls, alarms, and indications associated with the Instrument and Service Air Systems, including:

1. Location and function of component and/or system operating controls and control stations

2. Alarming locations and response to major system and component alarms

3. Plant, system, and component conditions or permissives required for local operation

4. Setpoints associated with major system alarms and/or interlocks (052.06.LP0338.008)

EXPLANATION:

A. CORRECT: This is correct, the valves auto open at 85 psig to supply station air to instrument air.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 131 of 199 B. INCORRECT: The valves will automatically open. Plausible if the student has a misconception concerning where the leak is an how the system will react to it, i.e., the instrument air and station air system cross-connect shutting to separate the systems.

C. INCORRECT: The valves will automatically open. Plausible if the student has a misconception concerning where the leak is an how the system will react to it, i.e., the instrument air and station air system cross-connect shutting to separate the systems.

D. INCORRECT: The valves will automatically open. Plausible if the student has a misconception concerning where the leak is an how the system will react to it, i.e., the instrument air and station air system cross-connect shutting to separate the systems.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 132 of 199 QUESTION # 066 Given the following:

• The mid shift is at minimum crew composition per OM 3.1, Operations Shift Staffing Rrequirements

• The STA holds an inactive SRO license

• At 0130 the 4th reactor operator leaves the site due to an emergency at home What action, if any, is required to restore shift staffing?

A. An emergency call out must begin immediately to restore shift staffing.

B. The OS2 must turn over command and control and assume the duties of the 4th license.

C. The STA must report to the control room and remain there until the next crew arrives at 0650.

D. No action is required as long as the licensed operators assigned to each unit remain in the control room until the next crew arrives at 0650.

ANSWER:

A.

REFERENCE:

OM 3.1, Operations Shift Staffing Requirements, Rev 17 BANK FUNDAMENTAL K/A: G2.1.5: Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

LEARNING OBJECTIVE:

Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc. (PBN.PB.2.01.05)

STATE, in specific terms, the specifications associated with each of the following

1. Administrative Requirements.

2. unit Staff Qualifications

3. Procedures

4. High Radiation Area (057301LP3347.005)

EXPLANATION:

A. CORRECT: This is the correct answer. (OM 3.1, 5.4.1)

B. INCORRECT: This is not correct, an OS will assume the responsibility and authority of the Shift Manager (until relieved by a qualified individual), if the SM shift staffing requirement cannot be met or the SM is incapacitated. Plausible is the student applies this rule to the 4th RO position.

C. INCORRECT: This is only correct if SM or OS shift staffing requirements cannot be met, the STA shall be summoned to the Control room, and shall remain there until qualified replacement

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 133 of 199 personnel arrive. Plausible if the student applies this rule to the 4th RO position.

D. INCORRECT: This is not correct. Plausible is the student assumes the OM 31 requirements are more restrictive than the tech spec requirements.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 134 of 199 QUESTION # 067 Given the following:

• A Unit 1 Core reload is in progress per RP-1C, Refueling

• You are the Unit 1 Reactor Operator Which of the following conditions would require the refueling operation to cease?

A. RCS boron concentration has lowered from 2430 to 2375 PPM.

B. RHR Heat Exchanger inlet temperature has risen from 90°F to 110°F.

C. Source Range channel N-31 indicates a rise above the baseline count rate by a factor of 3.5.

D. Source Range channels N-32 AND N-40 indicate a rise above the baseline count rate by a factor of 1.5.

ANSWER:

C.

REFERENCE:

3 RP 1C, Refueling, Rev 74 BANK: 2007 PBNP NRC Exam FUNDAMENTAL K/A: G2.1.36 Conduct of Operations: Knowledge of procedures and limitations involved in core alterations.

LEARNING OBJECTIVE:

Knowledge of procedures and limitations involved in core alterations. (SD 86.1 2.1.36)

EXPLANATION:

A. INCORRECT: Boron concentration change of 100 PPM requires suspension of refueling operations. (P&L 3.2.14.a). Plausible because this is an unexpected change in boron concentration.

B. INCORRECT: Inlet temperature rise should be investigated but does not require suspension of refueling. Plausible because this is an unexpected change.

C. CORRECT: One channel rising by a factor of three or more requires suspension of refueling. (P&L 3.2.14.b)

D. INCORRECT: Two channels rising by a factor of TWO or more requires suspension. (P&L 3.2.14.c). Plausible because this is one of the precautions, but not to the level required to cease refueling.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 135 of 199 QUESTION # 068 Given the following:

• Unit 1 is at Rated Thermal Power

• Pressurizer Level Channel Selector Switch is in OPERATE position

• Letdown isolates due to a Pressurizer level instrument failure

• The Control Operator states that the controlling level channel has failed Which of the following would validate the accuracy of the operator's diagnosis that the controlling channel, and not the backup channel, has failed?

A. Backup heaters turn off B. Letdown orifice valves close C. The speed of the Charging Pump in AUTO is rising D. 1C04 1C 1-3, PRESSURIZER LEVEL HIGH OR LOW is LIT ANSWER:

C.

REFERENCE:

883D195 Sh18, Logic Diagram Pressurizer Pressure and Level Control, Rev 13 FUNDAMENTAL BANK: 2012 Diablo Canyon NRC Exam K/A: G2.1.45 Ability to identify and interpret diverse indications to validate the response of another indication.

LEARNING OBJECTIVE:

Ability to identify and interpret diverse indications to validate the response of another indication. SD 86.1.2.1.45 EXPLANATION:

A. INCORRECT: Both the controlling and backup channel will turn off heaters. Plausible, if the student has the misconception that controlling does not have this interlock also.

B. INCORRECT: Both the controlling and backup channel will close the letdown orifice valves. Plausible, if the student has the misconception that controlling does not have this interlock also.

C. CORRECT: This is correct, only the controlling channel has a direct input to the speed signal of the charging pumps. If the backup channel failed causing letdown to isolate, the charging pump speed would slow, based on actual level rising caused by the charging \

letdown flow mismatch.

D. INCORRECT: Both the controlling and backup channel will cause this annunciator to alarm. Plausible, if the student has the misconception that controlling does not have this interlock also.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 136 of 199 QUESTION # 069 See the PROVIDED information from the ROD 3.2, ICRR for a Unit 1 Reactor Startup.

Which of the following actions are required to be perfomed per OP 1B, Reactor Startup?

A. Perform a 100 pcm control rod withdraw.

B. Perform the next 400 pcm control rod withdraw.

C. Notify the Shift Manager and insert control bank rods.

D. Immediately trip the reactor and notify the Shift Manager.

ANSWER:

B.

REFERENCE:

OP 1B, Reactor Startup, Rev 69 BANK HIGHER K/A: G2.2.1 Equipment Control: Ability to perform pre-startup procedures for the facility, including operating those controls associated with plant equipment that could affect reactivity.

Proposed reference to be provided to the applicants during examination:

Modified U1C35 Rod 3.2 page 2 Burnup Range (3000-4000)

LEARNING OBJECTIVE:

EVALUATE ECP and ICRR data and DETERMINE if critical rod height will satisfy Technical Specifications and Administrative Guidelines.

(055.01.LP0183.006)

EXPLANATION:

A. INCORRECT: A 100 PCM pull is not required by OP 1B at this time.

Plausible if the student has a misconception of the requirements of OP 1B and the requirements of when <100 PCM rod withdraws are required.

B. CORRECT: A 400 PCM rod withdraw is correct.

C. INCORRECT: This is required if it is determined that criticality will be achieved prior to rod withdraw to the (-)750 PCM position, which is not true in this case. Plausible if the student has a misconception of this requirement.

D. INCORRECT: The conditions which require a reactor trip during start up are not met. Plausible is the student has a misconception of those requirements.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 137 of 199 QUESTION # 070 Given the following:

Unit 2 is being started up per OP 1B, Reactor Startup, following a mid-cycle trip Control Bank A has been withdrawn to 35 steps

'A' and B RCPs both trip due to a bus voltage fluctuation which quickly stabilizes back to normal voltage What actions are required to be taken?

Respond to the automatic reactor trip in accordance with EOP-0, Reactor Trip or Safety Injection.

Manually trip Unit 2 reactor and respond to manual trip in accordance with EOP-0, Reactor Trip or Safety Injection.

Abort the startup and drive control bank rods to 0 steps. When RCPs are restored, initiate a new copy of OP 1B and recommence startup.

Suspend the startup; enter AOP-1B, Reactor Coolant Pump Malfunction, to restart the RCPs and continue with the startup with the Shift Managers permission.

ANSWER:

B.

REFERENCE:

OP 1B, Reactor Startup, Rev 68 HIGHER BANK K/A: G2.2.2 Equipment Control: Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels.

Learning Objective:

ACCOUNT for, and DESCRIBE the response to, various alarms existing or received during a Reactor Startup. (055.01.LP0183.010)

EXPLANATION:

INCORRECT: This will not cause an automatic trip. The second portion is the correct procedure to continue in. Plausible is the student deduces that an automatic reactor trip will happen, as would at a higher power level.

CORRECT: This is the correct action as the reactor will not automatically trip, and also the correct procedure to transition to.

INCORRECT: The first portion is correct, the start up will be aborted, but the control rods will not be driven in. A new OP 1B will be performed once the fault is identified, and reactor coolant pumps are restored.

Plausible if the student applies the Delayed Restart Attempt portion of the procedure, as this ensures all control bank rods are inserted.

INCORRECT: This is not correct, even though there will not be an automatic trip, and AOP-1B will be transitioned to eventually to deal with the loss of the reactor coolant pumps, the plant is required to be tripped.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 138 of 199 Plausible if the student assumes a trip is not required, and transitions to AOP-1B to mitigate and correct the loss of the reactor coolant pumps.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 139 of 199 QUESTION # 071 Which of the following conditions defines MODE 3 per Technical Specifications?

Keff < 0.99 TAVE <350°F Keff > 0.99 TAVE >350°F Keff > 0.99 TAVE <350°F Keff < 0.99 TAVE >350°F ANSWER:

D.

REFERENCE:

TS 1.1, Definitions Table 1.1-1, Rev Amend 201/206 BANK: 2011 Comanche Peak NRC Exam FUNDAMENTAL K/A: G2.2.35 Equipment Control: Ability to determine Technical Specification Mode of Operation.

Learning Objective:

Ability to determine Technical Specification Mode of Operation (2.02.35057)

EXPLANATION:

INCORRECT: Plausible because Keff is correct, however temperature is not.

INCORRECT: Plausible because temperature is correct, however Keff is not.

INCORRECT: Plausible because Keff is correct for Mode 2 and temperature is correct for Mode 4.

CORRECT: When Keff is < 0.99 and Tave is > 350°F the unit is in MODE 3, HOT STANDBY.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 140 of 199 QUESTION # 072 Given the following:

You are a Relief Crew CO, performing a valve lineup in the PAB during a Refueling Outage The valve line up requires the you to go 10 into the overhead to operate a valve What action are you going to take, if any, per the Operations Routine Job RWP?

No action is required, the PAB is routinely surveyed, any hazards will be posted.

No action is required if less than a shift since the last survey was performed.

You are required to notify RP prior to taking any actions > 6ft in the overhead.

You are required to have a meter to check dose rates when going > 6 ft in the overhead.

ANSWER:

C.

REFERENCE:

3 PBN LOC 14D 004L, OM-OE, Rev 0 (CR01982315, Improper Radiation Worker Practices)

NEW FUNDAMENTAL K/A: G2.3.7 Radiation Control: Ability to comply with radiation work permit requirements during normal or abnormal conditions.

EXPLANATION:

INCORRECT: Action is required due to having to go above 6 ft.

Plausible if the student has a misconception that since this is a test, and a routine task, the valve area should have already been checked.

INCORRECT: The action needs to be reported to RP prior actions being taken. Plausible if the student has a misconception that the completion of a survey will satisfy this requirement CORRECT: Per OPS Standing RWP accessing any areas >6ft in the overhead requires prior notification to RP.

INCORRECT: This is not required, unless directed by RP, and to be directed by RP, they would have had to be notified. Plausible, if the student has a misconception of self monitoring and what can be done while self monitoring.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 141 of 199 QUESTION # 073 Prior to moving irradiated fuel assemblies with the Spent Fuel Bridge and the Spent Fuel Handling Tool, we ensure that long-handled tools have holes drilled in them.

Why do we require long handled tools to fill with water while working in the SFP?

To prevent radiation streaming from highly radioactive material stored under water.

Filling the tools with water helps minimize buoyancy for easier handling of fuel assemblies.

The water provides shielding which minimizes spurious Spent Fuel Pool Area Radiation Monitor Alarms.

Keeping the inside of the tool wet prevents airborne conditions from developing thus minimizing personnel exposure.

ANSWER:

A.

REFERENCE:

3 0-SOP-FH-001, Fuel/Insert/Component Movement in the Spent Fuel Pool New Fuel Vault, 3.4.1, Rev 24 BANK: 2012 PBNP NRC Exam (Last 2 NRC Exams)

FUNDAMENTAL K/A: G2.3.12 Radiation Control: Knowledge of radiological safety principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Learning Objective:

Describe the function, purpose, design basis, physical location, and operating characteristics of the Spent Fuel pool Equipment.

(112.01.LP0260.001)

EXPLANATION:

CORRECT: The flooding prevents radiation streaming from highly radioactive material stored under water. (P&L 3.4.1)

INCORRECT: Buoyancy is not an issue when dealing with fuel assembly weight. Plausible is the student does not know that the weight calculations for the overloads takes into consideration a flooded handle.

INCORRECT: Streaming radiation from the tool should not effect an area radiation monitors, but should be a concern for the operator. Plausible is the student assumes that a hollow tool will cause the radiation in the area to increase due to the lack of shielding in the tool.

INCORRECT: This is a common practice to minimize airborne radiation, but would only be an issue is the handle was hollow and the top was open. Plausible is the student believes the construction of the tool to be such that loose surface contamination is an issue when working with the tool.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 142 of 199 QUESTION # 074 Many of PBNPs Emergency Operating Procedures (EOPs) contain the following note: "RCPs should be run in order of priority to provide normal PZR spray . . ."

Concerning this note, which choice completes the following statement:

The correct starting order of the pumps is (1) and the bases is that with the plant design of one surge line and two spray lines, the RCP in the loop (2) the pressurizer surge line, if running, provides normal pressurizer spray.

(1) (2)

A then B with A then B opposite B then A with B then A opposite ANSWER:

C.

REFERENCE:

EOP-1.2, Post LOCA Cooldown and Depressurization, Rev 31 BG-EOP-1.2, Background Document for Post LOCA Cooldown and Depressurization, Rev 26 NEW FUNDAMENTAL K/A: G2.4.20 Emergency Procedures / Plan: Knowledge of the operational implications of EOP warnings, cautions, and notes.

Learning Objective:

Knowledge of operational implications of EOP warnings, cautions and notes. (2.04.20)

EXPLANATION:

The note is contained in the following procedures:

EOP-0.1 EOP-1.2 ECA-1.3 EOP-0.2 EOP-3.1 ECA-2.1 EOP-0.3 EOP-3.2 ECA-3.1 EOP-0.4 EOP-3.3 ECA-3.2 EOP-1.1 INCORRECT: The pump starting configuration is wrong but purpose of the note, is correct. Plausible if the student has a misconception of the starting sequence of the pumps and the design of the plant, specifically which loop has the surge line.

INCORRECT: The pump starting configuration is wrong along with the purpose of the note. Plausible if the student has a misconception of the starting sequence of the pumps.

CORRECT: This is the correct starting configuration as well as meaning

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 143 of 199 of the note, to run one pump to get normal/near normal pressurizer spray flow. Running the pump in the loop with the surge line, will cause normal pressurizer spray using only one RCP. (BG EOP-1.2 step 10 note 1)

INCORRECT: The pump starting configuration is correct but purpose of the note, incorrect. Plausible if the student has a misconception of the starting sequence of the pumps and the design of the plant, specifically which loop has the surge line.

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 144 of 199 QUESTION # 075 Given the following:

Unit 2 is at Rated Thermal Power The following events happen:

2105 Power is lost to all Control Room annunciators 2120 The Shift Manager (SM) declares an Unusual Event 2125 The NARS form is completed, and the SM directs an operator to make the notification to the state and counties 2130 A reactor trip occurs on Unit 1 and the SM declares a Site Area Emergency 2135 The SM completes a new Emergency Notification Form and directs the notification of state and counties Which of the following indicates the latest time when the initial state and county notifications must be completed?

2135 2140 2145 2150 ANSWER:

A.

REFERENCE:

EPIP 1.2, Emergency Classification, Rev 52 BANK: 2009 McGuire NRC Exam FUNDAMENTAL K/A: G2.4.39 Emergency Procedures / Plan: Knowledge of RO responsibilities in emergency plan implementation.

Learning Objective:

Knowledge of RO responsibilities in emergency plan implementation (043.03.LP2000.006)

EXPLANATION:

CORRECT: The original (initial) declaration time was 2120. State and counties are required to be notified within 15 minutes of the initial emergency declaration for the UNUSUAL EVENT. This means the state and counties must be notified by 2135.

INCORRECT: See Above. Plausible if the student has the misconception that the time requirement states when the Emergency Notification Form has been completed.

INCORRECT: See Above. Plausible if the student has the misconception that if a change in declaration level occurs prior to notification, the notification clock resets to the time of the new declaration.

INCORRECT: See Above. Plausible if the student has the misconception that the initial notification clock resets and then starts

2015 Point Beach Nuclear Plant Reactor Operator Initial License Exam Page 145 of 199 when the Emergency Notification Form is complete.

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 146 of 199 QUESTION # 076 Given the following:

A loss of Coolant Accident (LOCA) occurred from 100% power on Unit 1.

The crew is performing the actions of EOP-0, Reactor Trip or Safety Injection currently at Step 21, Check if RCPs Should be Stopped.

Tcold is 315°F and Lowering.

RCS pressure is 90 psig and Lowering.

Containment pressure has peaked at 10 psig and is Stable.

All Engineered Safeguards Equipment functions as designed with the exception of Safety Injection Pumps which cannot be started.

Reactor Coolant Pumps (RCPs) are running.

What actions are required regarding the RCPs AND what procedure transition, if any, is required?

Trip RCPs AND continue in EOP-0, Reactor Trip or Safety Injection.

Trip RCPs AND transition to CSP-P.2, Response to Anticipated Pressurized Thermal Shock Condition Keep RCPs running AND continue in EOP-0, Reactor Trip or Safety Injection.

Keep RCPs running AND transition to CSP-P.2, Response to Anticipated Pressurized Thermal Shock Condition.

ANSWER:

C.

REFERENCE:

EOP-0, Reactor Trip or Safety Injection Unit 1, rv. 59 CSP-ST.0, Critical Safety Function Status Trees Unit 1, rv. 8 OM 3.7, AOP and EOP Procedure Usage for Response to Plant Transients, rv. 23 BANK: 2010 Beaver Valley 2 HIGHER K/A: 011EA2.01 Large Break LOCA: Ability to determine or interpret the following as they apply to a Large Break LOCA: Actions to be taken, based on RCS temperature and pressure - saturated and superheated.

LEARNING OBJECTIVE:

STATE the basis for Reactor Coolant Pump trip criteria and LIST the conditions when Reactor Coolant Pump Trip Criteria does not apply.

(031.02.LP0405.009)

DISCUSS the prioritization of the Core Cooling and Inventory CSFs including color coding, relationship to other CSFs, and relationship to the EOPs and ECAs. (043.03.LP1997.009)

EXPLANATION:

Incorrect. RCPs are not required to be tripped based on plant conditions.

Plausible that they are tripped because during large break LOCAs, the operation of the RCPs has little if any effect during mitigation and Tuesday, December 23, 2014 146

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 147 of 199 recovery. Remaining in EOP-0 is the correct procedural application.

Incorrect. RCPs are not required to be tripped based on plant conditions.

It is plausible that a transition to CSP-P.2 is made since Tcold is < 340

°F; however, EOP rules of usage specify that CSFST monitoring should not occur until directed by EOP-0 or a transition is made to another procedure. Neither of these has occurred based on stated plant conditions. Additionally, the SRO candidate should have the knowledge that this is a yellow path CSF, which requires transition based on his judgment only.

Correct. RCPs are required to be running during these plant conditions.

Although RCS subcooling is approximately 17 °F, which meets trip criteria, since there is no SI flow, RCPs are NOT secured. The SRO must be able to assess plant conditions and prescribe procedures to proceed based on these conditions. Remaining in EOP-0 is the correct course of action until a transition to EOP-1 is directed.

Incorrect. Correct that RCPs must be kept running, CSP-P.2 transition criteria is met, however, not required based on rules of usage.

Tuesday, December 23, 2014 147

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 148 of 199 QUESTION # 077 Given the following:

Unit 1 is at Rated Thermal Power Unit 2 is in MODE 5 The following issues happen to the Component Cooling Water (CCW)

System 1 February at 0600, HX-12B, CCW Heat Exchanger develops a tube leak and is declared inoperable 2 February at 0200, 1P-11A, CCW Pump is declared inoperable due to vibrations 4 February at 1400, HX-12C, CCW Heat Exchanger is declared inoperable due to a tubesheet leak 4 February at 2200, 1P-11A is repaired, tested and declared OPERABLE 6 February at 2100, 1P-11B, CCW Pump is declared inoperable for a bearing failure 7 February at 1300, HX-12B is repaired, tested and declared OPERABLE Assuming that the current inoperable components cannot be repaired and brought to an OPERABLE status, which describes the latest allowable date and time which Unit 1 must be in MODE 3? (See provided reference) 7 February at 2000 8 February at 0800 10 February at 0300 10 February at 2000 ANSWER B.

REFERENCE:

TS 3.7.7, Component Cooling Water (CC) System rev 201 TS B 3.7.7, Component Cooling Water (CC) System Bases rev 201 Proposed reference to be provided to the applicants during examination:

TS 3.7.7, Component Cooling Water (CC) System NEW HIGHER K/A: 026G2.2.38 - Loss of Component Cooling Water: Knowledge of conditions and limitations in the facility license.

LEARNING OBJECTIVE:

Given specific plant conditions, assess and apply Technical Specification Requirements, as appropriate (057.02.LP3343.002)

EXPLANATION:

INCORRECT: This answer is based on the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (+6) for HX-12C.

This is incorrect as when HX-12B is declared OPERABLE (at the 71 hour8.217593e-4 days <br />0.0197 hours <br />1.173942e-4 weeks <br />2.70155e-5 months <br /> point for HX-12C), that ACTION CONDITION is exited, so HX-12C does not need to be OPERABLE to meet the LCO. Plausible if the student does not realize the ACTION CONDITION is exited with the repair of HX-Tuesday, December 23, 2014 148

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 149 of 199 12B.

CORRECT: This answer is based on the 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> (+6) from not meeting the LCO. The LCO is not met with the inoperable declaration of the 1P-11A pump, the LCO is met with the inoperable declaration of the HX-12B, as the system design and bases allows for one heat exchanger to be inoperable, and still met the LCO. Once the LCO is not met, it must be met within 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br />.

INCORRECT: This answer is based on 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> (+6) for 1P-11B. This is incorrect, this would be longer than the 144 hour0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> requirement to met the LCO. Plausible due to 1P-11B is the only component which is inoperable and does not cause the LCO not to be met.

INCORRECT: This answer is based on the 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> (+6) for HX-12C.

This is incorrect, as the clock actually started when 1P-11A was declared inoperable. Plausible if the student applies the 144 hour0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br /> requirement to components which are not considered OPERABLE.

Tuesday, December 23, 2014 149

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 150 of 199 QUESTION # 078 Given the following:

Unit 1 is at Rated Thermal Power A steam line break develops inside containment The conditions immediately following the steam line break are:

Reactor is tripped Containment pressure is 7 psig and Rising Pressurizer Level is 30% and Lowering Pressurizer Pressure is 1745 psig and Lowering Rapidly Mitigating the event, the crew has transitioned to and completed the actions required in EOP-2, Faulted Steam Generator Isolation Current plant conditions are:

Containment pressure is 26 psig (maximum) and Lowering Pressurizer Level is 22% and Rising RCS Pressure is 1735 psig and Slowing Rising RCS Subcooling is 100°F and Rising Which of the following identifies:

(1) Using the conditions immediately following the steam line break, which SI initiation signal was generated when 2 pressure bistables reached the setpoint AND (2) when transitioning from EOP-2, to which procedure should the SRO transition?

(1) Low Pressurizer Pressure (2) EOP-1.1, SI Termination (1) Low Pressurizer Pressure (2) EOP-1, Loss of Reactor or Secondary Coolant (1) High Containment Pressure (2) EOP-1.1, SI Termination (1) High Containment Pressure (2) EOP-1, Loss of Reactor or Secondary Coolant ANSWER D.

REFERENCE:

883D195 Sh 7 Rev 25 Safeguards Actuation Signals STPT 2.1 Rev 5 Safety Injection EOP-1, Loss of Reactor or Secondary Coolant Rev 42 EOP-2, Faulted Steam Generator Isolation Rev 24 Tuesday, December 23, 2014 150

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 151 of 199 MODIFIED: 2010 Surry NRC Exam HIGHER K/A: 040A2.04: Steam Line Rupture - Excessive Heat Transfer: Ability to determine and interpret the following as they apply to the Steam Line Rupture: Conditions requiring ESFAS initiation.

LEARNING OBJECTIVE:

IDENTIFY and DESCRIBE the controls, alarms, and indications associated with Safety Injection Actuation, including:

Location and function of component and/or system operating controls and control stations Alarming locations and response to major system and component alarms Plant, system, and component conditions or permissives required for operation Setpoints associated with major system alarms and/or interlocks (053.06.LP0486.017)

DISCUSS the basis for the SI Termination/Reinitiation criteria and the SI reduction sequence as presented in the EOP set. (031.02.LP1829.002)

EXPLANATION:

INCORRECT: The first part is incorrect, Pressurizer pressure logic is 2 of 3 with a setpoint of 1735. The second part is incorrect, the conditions for SI termination are met, pressurizer level greater than 13%, subcooling greater than 35°F and RCS pressure greater than 1725 psig for a non-adverse containment, but the path of the EOP network will be to EOP-1.

Plausible is the student has the misconception that both the reactor trip and SI setpoints are rate compensated (only the reactor trip is), and if the students recalls the requirements to enter EOP-1.1, SI Termination from within the EOP network, i.e., EOP-0, instead of the flowpath from EOP-2 to EOP-1.

INCORRECT: The first part is incorrect, Pressurizer pressure logic is 2 of 3 with a setpoint of 1735. The second part is correct, the EOP network will transition to EOP-1 based on the current plant conditions and EOP network flowpath. Plausible is the student has the misconception that both the reactor trip and SI setpoints are rate compensated (only the reactor trip is).

INCORRECT: The first part is correct is 2 of 3 and an SI signal will be generated. The second part is incorrect, the conditions for SI termination are met, pressurizer level greater than 13%, subcooling greater than 35°F and RCS pressure greater than 1725 psig for a non-adverse containment, but the path of the EOP network will be to EOP-1.

Plausible if the students recalls the requirements to enter EOP-1.1, SI Termination from within the EOP network, i.e., EOP-0, instead of the flowpath from EOP-2 to EOP-1.

CORRECT: The first part is correct is 2 of 3 and an SI signal will be generated. The second part is correct, the EOP network will transition to EOP-1 based on the current plant conditions and EOP network flowpath.

Tuesday, December 23, 2014 151

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 152 of 199 Original Question:

Reactor power = 100%

A steam line break develops inside containment Current Conditions:

Reactor is tripped RCS Pressure is 1780 psia and decreasing Containment pressure = 17.7 psia and increasing Pzr Level = 25% decreasing Indicated RCS Subcooling (ICCM) = 100°F AFW flow to each intact Steam Generator is 250 gpm.

The team has reached step 8 of 1-E-2, CHECK IF SI FLOW SHOULD BE REDUCED Based on the above conditions, which one of the following states: (1) which SI initiation signal will be generated in only 2 pressure switches are at the setpoint and (2) when transitioning from 1-E-2, to which procedure should the SRO transition?

(1) RCS Pressure Low (2) 1-ES-1.1, SI TERMINATION (1) RCS Pressure Low (2) 1-E-1, LOSS OF REACTOR OR SECONDARY COOLANT (1) Hi CLS (2) 1-ES-1.1, SI TERMINATION (1) Hi CLS (2) 1-E-1, LOSS OF REACTOR OR SECONDARY COOLANT Tuesday, December 23, 2014 152

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 153 of 199 QUESTION # 079 Given the following:

Unit 1 is at Rated Thermal Power A Reactor Trip occurred coincident with a Loss of Offsite Power Only 1A-06 and 1B-04 are energized from its Emergency Diesel Generator Buses 1A-05, and 1B-03 are NOT energized The crew transitions to CSP-H.1, Response to Loss of Secondary Heat Sink, and is performing the first step The diesel generator output breaker supplying 1A-06 trips open for no apparent cause Which of the following identifies:

(1) the required action(s)

AND (2) the basis for the action(s)?

(1) Transition to ECA-0.0, Loss of All AC Power (2) because the ECA will direct actions to establish a heat sink with the TDAFW pump (1) Transition to ECA-0.0, Loss of All AC Power (2) because all CSPs are written on the premise that at least one AC safeguards train is energized.

(1) Remain in CSP-H.1, to recover and establish a heat sink with the TDAFW pump (2) because restoring heat sink is the highest priority evolution in progress (1) Remain in CSP-H.1 and perform AOP-19B, Train 'B' Safeguards Bus Restoration, in parallel to restore the breaker alignment and power to 1A-06 (2) because all CSPs are written on the premise that at least one AC safeguards train is energized.

ANSWER:

B

REFERENCE:

PBE-7033 Electrical Distribution Rev 6 CSP-H.1, Response to Loss of Secondary Heat Sink Rev 35 ECA-0.0, Loss of All AC Power Rev 61 BG-ECA-0.0, Background for Loss of All AC Power Rev 35 AOP-19B, Train "B" Safeguards Bus Restoration Rev 10 BANK: 2007 VC Summer NRC Exam HIGHER K/A: 056G2.4.21 - Loss of Off-site Power: Knowledge of the Tuesday, December 23, 2014 153

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 154 of 199 parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

LEARNING OBJECTIVE:

Given access to the Site Specific simulator, IMPLEMENT ECA-0.0 to restore AC Power to at least one Safeguards Bus (031.02.LP0462.006)

EXPLANATION:

INCORRECT: The transition to ECA-0.0 is the correct transition (on indication that both AC safeguards trains are deenergized. and ECA-0.0 will direct actions to establish AFW flow with the TDAFW pump thus establishing a heat sink, but this is not the reason for the transition.

Plausible because the transition is correct and the procedure will establish AFW flow.

CORRECT: The transition is correct, and the reason is also correct.

NCORRECT: Remaining in CSP-H.1 is not correct, the reason for the decision is also not correct. Plausible because of the importance of the RED path, and transitioning to CSP-H.1 prior to the loss of the safeguards bus.

INCORRECT: Remaining in CSP-H.1 is not correct, the reason for the decision is also not correct. Plausible because CSP-H.1 refers the crew to AOP-22, but AOP-19B, is more realistic to restore the bus. This will attempt to restore a safeguards train, and with a safeguards train energized, you remain in CSP-H.1 Tuesday, December 23, 2014 154

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 155 of 199 QUESTION # 080 Given the following:

Unit 1 and Unit 2 are at Rated Thermal Power K-2B, Instrument Air Compressor is out of service for repair A Service Water (SW) leak occurs in the plant K-2A, Instrument Air Compressor, AND K-3A, Service Air Compressor trip The crew enters AOP-9A, Service Water System Malfunction, and isolates the South SW header to isolate the SW leak Complete the following statement:

K-3B will (1) , an OS will enter AOP-5B, Loss of Instrument Air and (2) .

(1) (2) trip on high SW discharge remain in AOP-5B, no transitions temperature from AOP-5B are required continue to run, SW remain in AOP-5B, no transitions cooling is not required from AOP-5B are required trip on high SW discharge trip both reactors and enter temperature EOP-0, Reactor Trip or Safety Injection, while concurrently performing AOP-5B, Loss of Instrument Air continue to run, SW trip both reactors and enter cooling is not required EOP-0, Reactor Trip or Safety Injection, while concurrently performing AOP-5B, Loss of Instrument Air ANSWER:

C.

REFERENCE:

AOP-5B, Loss of Instrument Air Rev 41 M-207 Sh 1, P&ID Service Water Rev 84 M-207 Sh 1A, P&ID Service Water Rev 40 FSAR Section 9.7 Instrument Air and Service Air LP0338 Instrument and Service Air Rev 14 MODIFIED: 2011 Callaway NRC Exam HIGHER K/A: 065AA2.04: Ability to determine and interpret the following as they apply to the Loss of Instrument Air: Typical conditions which could cause a compressor trip (e.g., high temperature).

LEARNING OBJECTIVE:

RECOGNIZE and ANALYZE the effects of a loss of Service Water Tuesday, December 23, 2014 155

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 156 of 199 System on:

Instrument Air Compressors Diesel Generator Coolers (G01 and G02)

Auxiliary Feedwater Pumps Component Cooling Water Heat Exchangers e. New Battery and Inverter Room Coolers (051.06.LP0086.007)

DISCUSS how the following conditions/events could affect overall operation of the plant.

Loss of Instrument Air Gland Steam Condenser Tube Leak c. Loss of Condenser Vacuum Abnormal conditions for Reactor Coolant Pump vibration and/or oil, as indicated by alarms Steam binding or overheating of Auxiliary Feedwater Pumps Feedwater Heater and/or Heater Drain Tank malfunction Loss of DC Control Power to Main Generator Exciter Breakers h. Loss of or malfunction of the CVCS System. (069.04.LP0352.005)

EXPLANATION:

INCORRECT: The compressor having lost the required SW cooling will trip on high SW discharge temperature. The direction is incorrect, AOP-5B will also have you trip the reactor based on a loss of all compressed air. Plausible because there are actions taken in AOP-5B to mitigate a loss of instrument air and AOP-5B will be performed in parallel with EOP-0.

INCORRECT: The status is incorrect K-3B is the SW cooled air compressor, and AOP-5B will also have you trip the reactor based on a loss of all compressed air. Plausible because one of the two service air compressors has a SW discharge temperature trip and the mitigate a loss of instrument air and AOP-5B will be performed in parallel with EOP-0.

CORRECT: The compressor having lost the required SW cooling will trip on high SW discharge temperature this results in a loss of instrument and service air compressors. OS1 will enter AOP-5B, transition to EOP-0, trip and stabilize the plant and branch back to AOP-5B, with the purpose of performing both in parallel.

INCORRECT: The status is incorrect K-3B is the SW cooled air compressor, OS1 will enter AOP-5B, transition to EOP-0, trip and stabilize the plant and branch back to AOP-5B, with the purpose of performing both in parallel. Plausible because one of the two service air compressors has a SW discharge temperature trip and the other does not.

Tuesday, December 23, 2014 156

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 157 of 199 Original Question Given the following conditions:

Reactor Power is 100%.

CKA01A, A Air Compressor, is out of service for overhaul.

CKA01B, B Air Compressor, trips due to low oil pressure.

The in service Central Chiller, SGB01A, trips and the Standby Central Chiller, SGB01B, can not be started.

Which of the following directions will the Control Room Supervisor (CRS) give to the Crew AND what is the status of CKA01C, C Air Compressor?

CRS DIRECTION CKA01C STATUS Trip the Reactor and enter E-0, Will continue to run with ESW Reactor Trip or Safety Injection providing cooling water flow Commence a Turbine Load Will trip on high air discharge Reduction to 90% IAW OTO-MA- temperature 00008, Rapid Load Reduction Commence a Turbine Load Will continue to run with ESW Reduction to 90% IAW OTOMA- providing cooling water flow 00008, Rapid Load Reduction Trip the Reactor and enter E-0, Will trip on high air discharge Reactor Trip or Safety Injection temperature Tuesday, December 23, 2014 157

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 158 of 199 QUESTION # 081 Given the following:

American Transmission Company (ATC) notified PBNP that a Solar Magnetic Disturbance (SMD) warning of Kp8 has been issued The crew has implemented AOP-31, Solar Magnetic Disturbance Alert

Response

Generator MVARS are in the OUT direction Attachment B, GSU Power Transformer DC Neutral Ground Current Data has been started for both 1X01 and 2X01, Main Transformers Which of the following actions should you direct per AOP-31? Enter AOP-17A, Rapid Power Reduction and . . . (See provided reference) only reduce load on Unit 1 to 78%.

only reduce load on Unit 2 to 63%.

reduce load in both Units to 78%.

reduce Unit 1 load to 78% and Unit 2 load to 63%.

ANSWER:

D.

REFERENCE:

AOP-31, Solar Magnetic Disturbance Alter Response Rev 1 Proposed reference to be provided to the applicants during examination:

AOP-31 AND Completed Logs (Attachment B) for 1X-01 and 2X-01 NEW HIGHER K/A: 077G2.4.47: Generator Voltage and Electric Grid Disturbances:

Ability to diagnose and recognize trends in an accurate and timely manner utilizing the appropriate control room reference material.

LEARNING OBJECTIVE:

DESCRIBE the procedures which govern operation of the electrical portion of the Main Generator. Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed or Non Licensed Operators. (052.04.LP0076.007)

EXPLANATION:

INCORRECT: Plausible if the student enters the RNO and takes actions only for the first unit meeting the required conations. i.e., If the students enters the RNO at step 5.c, unit 1 will meet this condition.

INCORRECT: Plausible if the student misreads the RNO step, or wants to first deal with unit which requires the larger downpower.

INCORRECT: Plausible if the student misinterprets the step and reduces power in both units to 78% because both units DC Neutral Current readings will be greater than or equal to 60.

CORRECT: This is the correct actions as required by AOP-31, step 5.

Tuesday, December 23, 2014 158

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 159 of 199 QUESTION # 082 Given the following:

Unit 1 is entering a refueling outage RCS temperature is 130°F and stable RCPs are secured Both trains of RHR are aligned for decay heat removal with 'A' train providing cooling Draindown to 40% Pressurizer level per OP 4D Part 1, Draining the Reactor Coolant System, was just completed The Control Operator then reports the following information to you:

Pressurizer level is slowly rising Source range counts are rising on both N-31 and N-32 1C04 1C 2-8 BA FLOW DEVIATION OR POTENTIAL DILUTION IN PROGRESS alarm is LIT As the Operations Supervisor of that unit what actions are you going to take?

Use OP 4A, Filling and Venting Reactor Coolant System, Attachment D, Blending Operation for Unit 1 and Unit 2, to borate the RCS from the blender using an available Charging Pump.

Enter AOP-6F, Low Concentration Water Pockets in RCS, and borate the RCS from the RWST using a Safety Injection Pump.

Enter SEP-1, Degraded RHR System Capability, in an effort to raise RHR flow to 3000 gpm ensuring prompt mixing of the RCS.

Enter CSP-S.1, Response to Nuclear Power Generation/ATWS, and emergency borate the RCS from the RWST using all available Charging Pumps.

ANSWER:

B.

REFERENCE:

AOP-6F, Low Concentration Water Pockets in RCS Rev 3 ARB 1C04 1C 2-8 BA Flow Deviation or Potential Dilution in Progress Rev 9 BANK: 2011 PBNB NRC Exam (Past 2 NRC Exams)

HIGHER K/A: 024AA2.06: Ability to determine and interpret the following as they apply to the Emergency Boration: When boron dilution is taking place.

LEARNING OBJECTIVE:

Given access to the Site specific Simulator or plant conditions, DIAGNOSE and RESPOND to CVCS malfunction in accordance with the appropriate procedures(s). (055.03.LP3718.004)

EXPLANATION:

INCORRECT: This is the proper procedure used to borate the RCS for recovery from refueling. Plausible as this is a procedure used during refueling.

CORRECT: Proper procedure and multiple entry conditions and is Tuesday, December 23, 2014 159

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 160 of 199 referenced in the ARB.

INCORRECT: This procedure would potentially result in a faster mixing of the boron, also the requirements to enter this procedure are not met.

Plausible as this procedure does provide a timely mitigation strategy.

INCORRECT: This procedure is used for boration, but when the plant is above 350°F. Plausible as this procedure is the one the students usually use for boration when in the EOP Network.

Tuesday, December 23, 2014 160

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 161 of 199 QUESTION # 083 Given the following:

Refueling operations are in progress on Unit 2 It has been 7 days since shutdown The Containment Equipment hatch is removed Irradiated fuel is being moved with the manipulator from the core to the containment upender for transfer to the Spent Fuel Pool A leak is reported from the refueling cavity drain line cleanout flange Refueling Cavity water level is slowly lowering Identify the direction the Core Load Supervisor should provide during this event.

Store the irradiated fuel assembly in any available core location, THEN shut the Transfer Tube Gate Valve.

Store the irradiated fuel assembly in the Spent Fuel Pool upender Frame Down, THEN shut the Spent Fuel Pool Transfer Canal Doors.

Store the irradiated fuel assembly in any available core location, THEN have the Control Room secure Unit 2 Purge Supply and Exhaust Fans.

Store the irradiated fuel assembly in the containment upender Frame Down, THEN shut the Transfer Tube Gate Valve AND the Spent Fuel Pool Transfer Canal Doors.

ANSWER:

A.

REFERENCE:

RP 1C, Refueling Rev 74 BANK: 2011 PBNB NRC Exam (Past 2 NRC Exams)

HIGHER K/A: 036G2.1.23 Fuel Handling Accident: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

LEARNING OBJECTIVE:

RECOGNIZE the radiological hazards associated with Refueling Operations. 112.01.LP0285.004)

RELATE water level relationships between the following:

Reactor Coolant System Upper Refueling Cavity Lower Refueling Cavity Transfer Canal (112.01.LP0285.005)

EXPLANATION:

CORRECT: This is an allowable storage location, and the correct action to keep the spent fuel pool from lowering.

INCORRECT: This is the correct location, but only if the transfer tube gate valve is shut. Plausible, based on the correct location, and if the student doesnt recall the requirement to close the transfer tube gate valve.

Tuesday, December 23, 2014 161

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 162 of 199 INCORRECT: This is an allowable storage location, if the equipment hatch was installed, which it is not. Plausible based on the allowable location.

INCORRECT: This is the incorrect place to store a fuel assembly with the leak being from the bottom of the lower cavity, storing the assembly in the containment upender will not provide enough protection from radiation. Plausible based on the students understanding of the spent fuel pool, and refueling cavity.

Tuesday, December 23, 2014 162

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 163 of 199 QUESTION # 084 Given the following:

A LOCA occurred on Unit 1 EOP-1, Loss of Reactor or Secondary Coolant is in progress The following plant conditions are noted:

Reactor Coolant System Pressure 460 psig Slowly Lowering Core Exit Thermocouples 470°F Slowly Lowering Containment pressure 23 psig Slowly Lowering Containment Sump 'B' level 88 inches Slowly Rising Containment Radiation Monitor 31 R/hr Stable Reactor Vessel Narrow Range Level 19 ft Slowly Rising Reactor Vessel Wide Range Level 45 ft Slowly Rising Reactor Coolant Pumps are secured Which CSP will the OS direct transition to?

CSP-C.1, Response to Inadequate Core Cooling CSP-C.2, Response to Degraded Core Cooling CSP-Z.1, Response to High Containment Pressure CSP-Z.2, Response to Containment Flooding ANSWER:

D.

REFERENCE:

CSP-ST.0, Critical Safety Function Status Trees Rev 8 MODIFIED: 2012 Comanche Peak NRC Exam HIGHER K/A: W/E15EA2.1 Containment Flooding: Ability to determine and interpret the following as they apply the (containment Flooding): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

LEARNING OBJECTIVE:

IMPLEMENT the Critical Safety Function Status Tree and Critical Safety Procedure rules of usage. (043.03.LP1995.013)

EXPLANATION:

INCORRECT: This is incorrect due to the narrow range reactor vessel level being above the setpoint for CSP-C.1, as well as temperature of the reactor core exit thermocouples being less than 700°F. Plausible as the student calculate RCS Subcooling and then must recall the status tree logic, so choices must be recalled and worked through.

INCORRECT: This is incorrect, wide range level requires one or two reactor coolant pumps to be running to be a valid reading. Plausible as the student must not only recall the requirements for wide and narrow range reactor vessel level, but the requirements for when each is valid.

INCORRECT: This in incorrect, because containment pressure is less than the setpoints. Plausible due to the student needing to recall the Tuesday, December 23, 2014 163

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 164 of 199 pressure requirements.

CORRECT: This is correct due to containment pressure being less than the setpoint, and sump level being higher than the setpoint.

Tuesday, December 23, 2014 164

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 165 of 199 Original Question Given the following conditions:

A Large Break Loss of Coolant Accident has occurred on Unit 2.

EOP-1.0B, Loss of Reactor or Secondary Coolant, is in progress.

The initial scan of Critical Safety Function Status Tree parameters indicate the following:

Pressurizer level is 0%.

Containment water level indicates 817 feet and slowly rising.

Containment Spray has automatically actuated and was verified in EOP-0.0B, Reactor Trip or Safety Injection.

Containment pressure is 14 psig and slowly lowering.

Containment pressure is 14 psig and slowly lowering.

Containment Radiation Monitors are in ALARM at 25 REM/hr.

Reactor Vessel Level Indication System has NO lights LIT.

Core Exit Thermocouples are 292°F.

Which procedure has entry priority to address the above conditions?

FRZ-0.2B, Response to Containment Flooding.

FRZ-0.1B, Response to High Containment Pressure.

FRI-0.2B, Response to Low Pressurizer Level.

FRZ-0.3B, Response to High Containment Radiation Level.

Tuesday, December 23, 2014 165

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 166 of 199 QUESTION # 085 Given the following:

Unit 1 was operating at Rated Thermal Power A rapid uncontrolled depressurization of both steam generators occurs The crew transitioned to CSP-P.1 Response to Imminent Pressurized Thermal Shock Condition The crew is assessing whether an RCS Temperature Soak is required Which of the following states the procedure requirement that determines if an RCS Temperature Soak is required, and the bases?

RCS Cooldown Bases for requirement Greater than 200°F Prevent flaw growth due to excessive tensile in any 60 minutes stresses on Reactor Vessel outer wall Greater than 100°F Prevent flaw growth due to excessive tensile in any 60 minutes stresses on Reactor Vessel inner wall.

Greater than 200°F Prevent flaw growth due to excessive tensile in any 60 minutes stresses on Reactor Vessel inner wall.

Greater than 100°F Prevent flaw growth due to excessive tensile in any 60 minutes stresses on Reactor Vessel outer wall ANSWER:

B.

REFERENCE:

CSP-P.1 Response to Imminent Pressurized Thermal Shock Condition, step 23 TS Bases 3.4.3 RCS Pressure and Temperature (P/T) Limits Rev 1-229-234 WOG ERG Background for FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, section 2.2 Analysis for Subsequent Cooldown following Imminent PTS Condition NEW FUNDAMENTAL K/A: W/E08G2.2.25 Pressurized Thermal Shock: Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

LEARNING OBJECTIVE:

COMPARE the operator-initiated recovery techniques assumed in the analysis to the actions in CSP-P.1 (043.03.LP1999.003)

EXPLANATION:

INCORRECT: Plausible because the student might incorrectly apply the 200°F per hour maximum allowable Pressurizer cooldown limit per TRM 3.4.2. In addition, the student can misunderstand the bases for limiting thermal stresses and their location.

CORRECT: TS Bases implement Appendix G limits in order to prevent Tuesday, December 23, 2014 166

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 167 of 199 the combined tensile stresses on the most limiting location (RV Inner Wall) from causing propagation of flaws.The FR-P.1 decision to conduct a soak is based on these restrictions.

INCORRECT: Plausible because the student might incorrectly apply the 200°F per hour maximum allowable Pressurizer cooldown limit per TRM 3.4.2. The type of stresses and their location is correct.

INCORRECT: The cooldown limit is correct. The second half is plausible if the student misunderstands the bases for limiting thermal stresses and their location.

Tuesday, December 23, 2014 167

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 168 of 199 QUESTION # 086 Given the following:

Unit 1 is being started up after a refueling outage The turbine is being synchronized with the grid The following annunciators are LIT 1C03 1D 2-1, 1P-1A OR B RCP LABYR SEAL DELTA-P LOW 1C03 1D 3-3, 1P-1B RCP NO. 1 SEAL WATER FLOW HIGH OR LOW The following indications are noted on the control boards 1P-1B RCP Labyrinth seal Delta-P (-)30H20 Stable 1P-1B RCP No. 1 Seal Outlet Temperature 132°F Slowly Rising RCP B Seal leakage High Range Flow 6 gpm Stable The OS1 will direct which of the following sequences of actions?

A. 1. Enter AOP-1A, Reactor Coolant Leak

2. Trip the Reactor

3. Transition to EOP-0, Reactor Trip or Safety Injection to stabilize the plant

4. Trip 'B' RCP

5. Shut 1RC-431B, PZR Normal Spray Valve

6. After 3 minutes, open 1CV-386, RCP Seal Water Bypass Control Valve B. 1. Enter AOP-1A, Reactor Coolant Leak

2. Transition to AOP-1B, Reactor Coolant Pump Malfunction

3. Trip 'B' RCP

4. Transition to EOP-0, Reactor Trip or Safety Injection

5. Trip the Reactor and to stabilize the plant

6. Shut 1RC-431 A AND 1RC-431B, PZR Normal Spray Valve

7. After 3 minutes, shut 1CV-270B, RCP B No. 1 Seal Water Return MOV C. 1. Enter AOP-1B, Reactor Coolant Pump Malfunction

2. Trip the Reactor

3. Transition to EOP-0, Reactor Trip or Safety Injection to stabilize the plant

4. Trip 'B' RCP

5. Shut 1RC-431B, PZR Normal Spray Valve

6. After 3 minutes, shut 1CV-270B, RCP B No. 1 Seal Water Return MOV D. 1. Enter AOP-1B, Reactor Coolant Pump Malfunction

2. Transition to EOP-0, Reactor Trip or Safety Injection

3. Trip the Reactor AND both RCPs concurrently

4. Stabilize the plant using EOP-0

5. Place steam dump mode selector in MANUAL

6. Shut 1RC-431B, PZR Normal Spray Valve

7. After 3 minutes, open 1CV-386, RCP Seal Water Bypass Control Valve Tuesday, December 23, 2014 168

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 169 of 199 ANSWER:

C.

REFERENCE:

AOP-1B, Reactor Coolant Pump Malfunction Rev 22 NEW HIGHER K/A: 003G2.4.45 Reactor Coolant Pump: Ability to prioritize and interpret the significance of each annunciator or alarm.

LEARNING OBJECTIVE:

Given access to the Site specific Simulator or plant conditions, RESPOND to the following events:

Rector Coolant System leakage Reactor Coolant Pump malfunctions Steam Generator Tube Leak Loose Parts Monitoring System alarm Rapid Power Reduction (055.03.LP2428.004)

EXPLANATION:

INCORRECT: This is the incorrect procedure to enter given the initial condition, the rest of the sequence is correct expect for the last step, this valve is checked shut. Plausible because AOP-1A will transition the operator to AOP-1B, and also the last step of the sequence may have been opened due to OP-4B, reactor coolant pump Operations.

INCORRECT: This is the incorrect procedure to enter given the initial condition, the rest of the sequence is correct except the pump is not tripped until after the plant is stabilized in EOP-0, and both spray valves are not shut. Plausible as this contains the called out transition to AOP-1B, where the transition in response A is implied given entry into AOP-1A, tripping the pump will stop further damage.

CORRECT: This is the correct procedure and sequence of events.

INCORRECT: This is the correct procedure to enter, but the sequence is incorrect. Plausible, because tripping the RCP can be done by one operator while the other is stabilizing the plant, and since there are no other failures the actions required to stabilize the plant are few. The last step of the squence open CV-386, can procedureally be accomplished by OP 4B, Reactor Coolant Pump Operations.

Tuesday, December 23, 2014 169

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 170 of 199 QUESTION # 087 Given the following:

Unit 1 is at 50% of Rated Thermal Power 1RC-431C, Pressurizer PORV has failed open 1RC-515, PRZ PORV Isolation MOV does NOT close when operated

1. Which of the following will be the indication that the PRT rupture disc has ruptured?

AND

2. What sequence of procedures would be required to mitigate this event?

EOP-0, Reactor Trip or Safety Injection EOP-0.1, Reactor Trip Response EOP-1, Loss of Reactor or Secondary Coolant EOP-1.1, SI Termination EOP-1.2, Post LOCA Cooldown and Depressurization

1. PRT temperature trending up.

2. EOP-0 to EOP-0.1

1. PRT temperature trending up.

2. EOP-0 to EOP-1 to EOP-1.1

1. Pressurizer relief valves outlet temperature trending down.

2. EOP-0 to EOP-0.1 to EOP-1.2

1. Pressurizer relief valves outlet temperature trending down.

2. EOP-0 to EOP-1 to EOP-1.2 ANSWER:

D.

REFERENCE:

EOP-0, Reactor Trip or Safety Injection Rev 59 EOP-1, Loss of Reactor or Secondary Coolant Rev 42 MODIFIED HIGHER K/A: 007A2.01 Pressurizer Relief/Quench Tank: Ability to (a) predict the impacts of the following malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Stuck-open PORV or code safety.

LEARNING OBJECTIVE:

DESCRIBE the function and/or purpose, design bases, interfaces with other systems, and operating characteristics of the RCS Vent System.

(069.02.LP0130.004)

EXPLANATION:

INCORRECT: The PRT temperature will trend down, and entry in to Tuesday, December 23, 2014 170

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 171 of 199 EOP-0.1 will not occur. Plausible as the temperature will rise until the rupture disc fails, then will lower based on the pressure of containment.

INCORRECT: The PRT temperature will trend down. SI termination will not happen with a stuck open PORV. Plausible as the temperature will rise until the rupture disc fails, then will lower based on the pressure of containment.

INCORRECT: The reduction in pressure would cause a reduction in temperature Psat/Tsat relationship when the rupture disc fails. Direct transition from EOP-0.1 to EOP-1.2 is not allowed. Plausible because the first part of the answer is correct.

CORRECT: The reduction in pressure would cause a reduction in temperature Psat/Tsat relationship when the rupture disc fails, the sequence is correct.

Tuesday, December 23, 2014 171

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 172 of 199 Original Question:

Plant conditions are as follows:

The unit is at 50% power.

Pressurizer PORV PCV-430 is stuck open.

Which of the following initially indicates that the PRT rupture disc has ruptured following the pressurizer PORV failing open?

and If the associated block valve, MOV-561 can NOT be shut, what sequence of procedures would be required?

1. PRT low level.

2. AP-RCS.1 to E-0 to ES-0.1

1. PRT temperature trending up.

2. E-0 to E-1 to ES-1.1

1. Pressurizer level trending down.

2. E-0 to ES-0.1 to ES-1.2

1. Pressurizer safety relief line temperature trending down.

2. E-0 to E-1 to ES-1.2 Tuesday, December 23, 2014 172

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 173 of 199 QUESTION # 088 Given the following:

A Large Break LOCA has occurred on Unit 1 Crew is currently in EOP-1.3, Transfer to Containment Sump Recirculation - Low Head Injection Crew is unable to open 1SI-850A, Train 'A' RHR Pump Suction From Containment Sump 'B', from the control room What direction should be provided to the crew to address this malfunction?

Continue in EOP-1.3 after directing the PAB AO to locally open 1SI-850A, in order to align 'A' train of sump recirc.

Direct the PAB AO to locally open 1SI-850A, in order to align 'A' train of sump recirc. Continue in EOP-1.3 only if this is successful.

Enter ECA-1.1, Loss of Containment Sump Recirculation, to perform the alternate alignment. Continue in parallel with actions in EOP-1.3 to recover A train of sump recirc.

The PAB AO will be directed to align RHR pump cross-connects in order to use the 'A' train of sump recirc; perform applicable sections of OP 7A, Placing Residual Heat Removal System in Operation, in parallel with EOP-1.3 to establish required lineup and continue with EOP-1.3.

ANSWER:

A.

REFERENCE:

EOP-1.3 U1, Transfer to Containment Sump Recirc - Low Head Injection Rev 52 OM 3.7 AOP AND EOP PROCEDURE USAGE Rev 23 MODIFIED: 2007 PBNP NRC Exam HIGHER K/A: 013G2.4.35: Engineered Safety Features Actuation System (ESFAS): Knowledge of local auxiliary operators tasks during an emergency and the resultant operational effects.

LEARNING OBJECTIVE:

Given appropriate conditions/parameters and access to the Site Specific Simulator, Implement the following procedures for the specified conditions:

c. EOP-1.3 to transfer to Containment Sump recirculation.

(031.02.LP0435.010)

EXPLANATION:

CORRECT: An attempt to locally open 1SI-850A is done per EOP-1.3.

The SRO will know that he must continue with efforts to establish a lineup using the other train in this situation. He applies the rules of usage for EOP procedure implementation for this situation.

INCORRECT: This is incorrect in execution of the EOP network.

Plausible if the student has a misconception of in the rules of usage in Tuesday, December 23, 2014 173

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 174 of 199 the EOP network, or if they mistakenly recall the step as a "do not continue until" type of step.

INCORRECT: Entry into ECA 1.1 is when both trains of sump recirc are lost. Plausible if the student mistakenly transitions there due to a perceived inability to line up one train of RHR.

INCORRECT: RHR pump cross-connects are used during Shutdown Cooling alignment, utilizing these cross-connects would not address the issue of being unable to open the 'A' train sump suction valve. Plausible if the students attempts the work around the problem of the failed valve.

Tuesday, December 23, 2014 174

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 175 of 199 Original Question:

Consider the following Unit 1 conditions:

A Large Break LOCA has occurred.

Crew is currently in EOP-1.3, "Transfer to Containment Sump Recirculation - Low Head Injection".

Crew is unable to open 1SI-850A, Train 'A' RHR pump suction from Containment sump 'B', from the control room or locally while performing step 17, Align RHR Sump Suction Valves.

What are the implications of this failure and what procedural actions should be taken?

'A' train of sump recirc will not be available; 'B' train will be used to mitigate the accident. Continue in EOP-1.3 and align 'B' train of sump recirc.

The RHR pump cross-connects must be opened in order to use the 'A' train of sump recirc; perform applicable sections of OP-7A, "Placing Residual Heat Removal System in Operation" in parallel with EOP-1.3 to establish required lineup and continue with EOP-1.3 1SI-850B, 'B' Train sump suction valve, will need to be aligned for 'A' train of sump recirc. Continue in EOP-1.3 and direct PAB AO to align alternate suction to 'A' recirc train.

1SI-850B, 'B' Train sump suction valve, will need to be aligned for 'A' train of sump recirc. Enter ECA-1.1, "Loss of Containment Sump Recirculation", to perform the alternate alignment.

Tuesday, December 23, 2014 175

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 176 of 199 QUESTION # 089 Given the following:

Unit 1 is operating at full power with operators responding to the following transient:

Containment pressure is 1.5 psig and Rising C01 B 1-4, UNIT 1 CONTAINMENT SUMP A LEVEL HIGH annunciator is LIT Steam Generator A level is 45% and Lowering Slowly Steam Generator B level is 64% and Stable Both Steam Generator pressures are approximately 800 psig and Stable The Operating Supervisor directs a manual Reactor Trip and Safety Injection Which of the following events is occurring and what is the required procedural flow path for the event?

EOP-0, Reactor Trip and Safety Injection EOP-1 Loss of Reactor or Secondary Coolant EOP-2, Faulted Steam Generator Isolation EOP-1.1, SI Termination Main Steam Line Break EOP-0, to EOP-2, to EOP-1, to EOP-1.1 Main Feed Line Break EOP-0, to EOP-2, to EOP-1, to EOP-1.1 Main Steam Line Break EOP-0, to EOP-2, to EOP-1.1 Main Feed Line Break EOP-0, to EOP-1, to EOP-1.1 ANSWER:

B.

REFERENCE:

EOP-0, Reactor Trip and Safety Injection Rev 59 EOP-1 Loss of Reactor or Secondary Coolant Rev Rev 42 EOP-2, Faulted Steam Generator Isolation Rev 24 NEW HIGHER K/A: 059A2.05: Ability to (a) predict the impacts of the following malfunctions or operations on the MFW; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Rupture in MFW suction or discharge line.

LEARNING OBJECTIVE:

Given access to Site Specific Simulator, IMPLEMENT the EOPs to respond to a faulted or ruptured Steam Generator. (031.02.LP0441.016)

EXPLANATION:

Tuesday, December 23, 2014 176

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 177 of 199 INCORRECT: Diagnosis is incorrect but plausible in that A steam generator water level is lowering due to loss of secondary coolant.

Second half is correct.

CORRECT: Diagnosis is correct. A steam generator pressure will not change significantly until the feed ring is uncovered, at which point the SG will blowdown similar to a steam line break. Second part is correct.

After the safety injection isolates main feedwater and the steam generator feed ring is uncovered, the lowering steam generator pressure will cause transition to EOP-2. Following isolation of the generator, the correct transition is to EOP-1, and then to EOP-1.1 for SI termination.

INCORRECT: Diagnosis is correct. Second part is plausible because the student may think that with stable SG pressure, there will be no transition to EOP-2 (and EOP-1 is Loss of Reactor or Secondary Coolant).

Incorrect because the SG will depressurize after the Feed line is isolated on Safety Injection and the feed ring is uncovered.

INCORRECT: Diagnosis is incorrect but plausible in that A steam generator water level is lowering due to loss of secondary coolant.

Second half is incorrect but plausible to believe that a transition directly from EOP-2 to EOP-1.1 is to be made once isolation of the Steam Generator occurs.

Tuesday, December 23, 2014 177

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 178 of 199 QUESTION # 090 Given the following:

Unit 1 Service Water overboard is isolated T-104A, A Waste Distillate Tank, is discharging to the Unit 2 Service Water overboard per OI 140B, Standard Radioactive Batch Liquid Release-Waste Distillate Tanks Prior to the discharge the RO recorded the following values for RE-223, Waste Distillate Tank Overboard Monitor on the discharge permit:

Background reading 2.67 E-5 µCi/cc Source check reading 1.42 E-4 µCi/cc Alert setpoint reading 7.67 E-4 µCi/cc Alarm setpoint reading 6.18 E-3 µCi/cc One hour after commencing the discharge the RO noted the following conditions:

RE-223, Waste Distillate Tank Overboard Monitor is reading 5.14 E-7

µCi/cc 2RE-229 SW Overboard Monitor Unit 2 is NOT in alarm What actions will the OS take?

(1) Determine TLCO 3.3.1 Instrumentation MET.

(2) The OS will direct the PAB AO to secure the discharge lineup and place the A Waste Distillate Tank on recirculation.

(3) Have Chemistry sample the A Waste Distillate Tank to validate the current discharge permit is good once RE-223 is repaired.

(1) Declare TLCO 3.3.1 Instrumentation NOT MET and enter TLCO 3.3.1.A and 3.3.1.B immediately.

(2) The OS will direct the PAB AO to secure the discharge lineup and place the A Waste Distillate Tank on recirculation.

(3) Chemistry will resample the A Waste Distillate Tank and issue a new discharge permit after RE-223 is repaired.

(1) Determine TLCO 3.3.1 Instrumentation MET.

(2) The OS will direct the PAB AO to bypass RE-223 and generate a work order to have RP to calibrate RE-223.

(3) Have Chemistry commence grab samples of the Unit 2 SW Overboard every hour until the A Waste Distillate Tank is discharged.

(1) Declare TLCO 3.3.1 Instrumentation NOT MET and enter TLCO 3.3.1.A and 3.3.1.B immediately.

(2) The OS will direct the PAB AO to bypass RE-223 and generate a work order to have RP calibrate RE-223.

(3) The discharge of the A Waste Distillate Tank can continue as long as 2RE-229 remains operable and Chemistry performs hourly grab samples from the A Waste Distillate Tank.

ANSWER:

Tuesday, December 23, 2014 178

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 179 of 199 B.

REFERENCE:

OI-140B, Standard Batch Release - Waste Distillate Tanks Rev 5 RAM 3.1.1, Restarting a Liquid Batch Release Rev 5 TRM 3.3.1 Instrumentation Rev 3 NEW HIGHER K/A: 073A2.03: Ability to (a) predict the impacts of the following malfunctions or operations on the Process Monitoring System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Calibration drift.

LEARNING OBJECTIVE:

Describe the procedures which govern operations of the Liquid Waste Disposal System. (051.04.LP0063.04)

EXPLANATION:

INCORRECT: Securing the discharge line up is correct and allowing the use of the current discharge permit is allowed in certain conditions. The TLCO being met is incorrect.

CORRECT: Per TLCO 3.3.1 Instrumentation the discharge must end due to the requirement to have RE-223 operable, there are no allowed back up monitors or grab samples allowed for the WDTs. Per OI 140B, P&L 3.8, a release point fail, shall be evaluated per RAM 3.1.1. Per RAM 3.3.1 P&L 3.2, states discharges secured shall have initial permit closed out, and restart will occur after sampling is completed and a new permit issued, once RE-223 is repaired.

INCORRECT: TLCO being met is incorrect. Grab samples are routinely allowed for many RMS monitors if they are OOS, especially when a second backup monitor is available.

INCORRECT: The TLCO being not met is correct and bypassing the monitor without securing the discharge is incorrect. Grab samples are routinely allowed for many RMS monitors if they are OOS, especially when a second backup monitor is available.

Tuesday, December 23, 2014 179

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 180 of 199 QUESTION # 091 Which of the following describes the applicability and Technical Specification Basis for the High Pressurizer Water Level reactor trip?

Modes 1 and 2; Provides a backup trip to PZR High Pressure reactor trip and ensures that water relief through the PZR safety valves will not occur Modes 1 and 2; Provides primary protection for loss of load events and ensures that on a PZR level channel failure, the PZR safety valves will not lift prior to the PZR High Pressure reactor trip.

Mode 1 above P-7; Provides a backup trip to PZR High Pressure reactor trip and ensures that water relief through the PZR safety valves will not occur.

Mode 1 above P-7; Provides primary protection for loss of load events and ensures that on a PZR level channel failure, the PZR safety valves will not lift prior to the PZR High Pressure reactor trip.

ANSWER:

C.

REFERENCE:

TS 3.3.1, RPS Instrumentation rev 1-239 TB B 3.3.1, Bases RPS Instrumentation rev 1-239 BANK: 2009 Wolf Creek NRC Exam FUNDAMENTAL K/A: 011G2.1.32. Pressurizer Level Control: Ability to explain and apply system limits and precautions.

LEARNING OBJECTIVE:

Discuss Technical Specification Definitions, Rules of Usage, Safety Limits and 1 Hour or less ACTIONS for Systems, Equipment, and Bases of LCOs and Safety Limits. (057.01.LP3366.017)

EXPLANATION:

INCORRECT: The MODE incorrect and bases is correct. Plausible because the basis is correct, for pressurizer pressure-high.

INCORRECT: The MODE and bases are incorrect. Plausible because the pressurizer high pressure trip actually provides this function, (as well as turbine trip-low autostop oil pressure) and the high water level trip is provided as a backup to pressurizer high pressure. The MODE is correct for pressurizer high pressure not water level high.

CORRECT: The applicability and bases are correct.

INCORRECT: The MODE is correct but the bases is incorrect. Plausible because the pressurizer high pressure trip actually provides this function, (as well as turbine trip-low autostop oil pressure) and the high water level trip is provided as a backup to pressurizer high pressure.

Tuesday, December 23, 2014 180

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 181 of 199 QUESTION # 092 Given the following:

A severe accident has occurred on Unit 1 The operators are currently implementing the Emergency Operating Procedures The following conditions exist:

The containment is adverse All Reactor Coolant Pumps are off Pressurizer level is off-scale low Narrow Range Reactor Vessel Level is 20 ft Wide Range RCS pressure is 400 psig Core exit thermocouples are reading 750°F Which of the following states the procedure transition that will be directed by the Operating Supervisor and the reason for the transition?

Transition to CSP-C.1, Response to Inadequate Core Cooling because core damage is occurring Transition to CSP-C.1, Response to Inadequate Core Cooling because core uncovery is occurring Transition to CSP-C.2 Response to Degraded Core Cooling because core damage is occurring Transition to CSP-C.2 Response to Degraded Core Cooling because core uncovery is occurring ANSWER:

D.

REFERENCE:

BG-CSP-C.2, Background Response to Degraded Core Cooling rev 5 CSP-C.2, Response to Degraded Core Cooling rev 8 BANK: 2012 Indian Point NRC Exam HIGHER K/A: 017A2.02 In-Core Temperature Monitor System (ITM): Ability to (a) predict the impacts of the following malfunctions or operations on the ITM system; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations:

Core damage.

LEARNING OBJECTIVE:

IMPLEMENT the CSPs to respond to plant conditions where the Core Cooling Status Tree is not satisfied. (043.03LP1997.011)

EXPLANATION:

INCORRECT: Plausible because CSP-ST.0 Critical Safety Function Status Trees uses 700 °F as one criterion (along with RVLIS indication) when evaluating if core damage is occurring. Because vessel level is above the level that indicates inadequate cooling, the transition to CSP-C.1 is not appropriate.

Tuesday, December 23, 2014 181

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 182 of 199 INCORRECT: Plausible because CSP-ST.0 Critical Safety Function Status Trees uses 700 °F as one criterion when evaluating if core damage is occurring. Although the superheated conditions indicate some core uncovery is occurring, the inadequate core cooling condition has not yet been reached so the transition to CSP-C.1 is not appropriate.

INCORRECT: Plausible because CSP-ST.0 Critical Safety Function Status Trees directs entry to CSP-ST.0 for this set of conditions. While the superheated conditions indicate some core uncovery is occurring, the inadequate core cooling condition has not yet been reached and the purpose of CSP-C.2 is to restore core cooling before an inadequate core cooling condition develops that would lead to core damage.

CORRECT: CSP-C.2 addresses conditions where core uncovery is likely. The actions in CSP-C.2 address restoration of inventory to prevent an inadequate core cooling condition.

Tuesday, December 23, 2014 182

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 183 of 199 QUESTION # 093 Given the following Units 1 and 2 are at Rated Thermal Power A loaded Dry Cask is being removed from the Spent Fuel Pool per RP 17, NUHOMS Dry Cask Loading and Storage, Part 4, Remove DSC/TC from Spent Fuel Pool During the critical lift, the rigging catastrophically fails, causing the Cask to fall back into the Spent Fuel Pool and come to rest on spent fuel racks which causes bubbles to rise from the fuel in the racks The dropped cask caused damage to the Spent Fuel Pool liner The Spent Fuel Pool water level is slowly lowering What procedural flow path will the OS1 direct to mitigate this event?

Transition to AOP-8F, Loss of Spent Fuel Pool Cooling.

Remain in RP 17, NUHOMS Dry Cask Loading and Storage, Part 4, Remove DSC/TC from Spent Fuel Pool, no transitions are necessary.

Transition to AOP-8C, Fuel Handling Accident in Primary Auxiliary Building, and when directed branch to AOP-8F, Loss of Spent Fuel Pool Cooling and perform in parallel.

Transition to AOP-8G, Dry Fuel Storage Cask Drop Tipover, and when directed branch to AOP-8C, Fuel Handling Accident in Primary Auxiliary Building and perform in parallel.

ANSWER:

C.

REFERENCE:

RP 17, Part 4 Remove DSC/TC from Spent Fuel Pool Rev 21 AOP-8C, Fuel Handling Accident in Primary Auxiliary Building Rev 1 NEW HIGHER K/A: 034A2.02 Fuel Handling Equipment: Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling System; and (b) based on those predictions, use procedures to correct, control or mitigate the consequences of those malfunctions or operations:

Dropped cask.

LEARNING OBJECTIVE:

DESCRIBE the procedures which govern Fuel Handling Operations.

Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed and Non-Licensed Operators (112.01.LP0285.003)

EXPLANATION:

INCORRECT: This is incorrect, RP 17 Part 4 directs immediate reference to AOP-8C in the event of a dropped cask, this will not mitigate the event of the dropped cask. Plausible due to meeting the entry conditions of AOP-8F.

INCORRECT: This is incorrect, RP 17 Part 4 directs immediate reference Tuesday, December 23, 2014 183

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 184 of 199 to AOP-8C in the event of a dropped cask, this procedure will not mitigate the event. Plausible due to already being in RP 17, and several other RP procedures contain compensatory actions to be taken in the event of a fuel handling accident.

CORRECT: RP 17 Part 4 directs immediate reference to AOP-8C.

AOP-8C directs parallel performance of AOP-8F if the conditions warrant as they do in this scenario.

INCORRECT: This is incorrect, the entry conditions AOP-8G are similar to the current conditions (cask lifting lugs failure or equipment failure resulting in a cask drop) but the purpose for the procedure is dealing with a cask drop outside of the primary auxiliary building. There is no directed transition to AOP-8C. Plausible if the student does not understand the required procedures for mitigating the event.

Tuesday, December 23, 2014 184

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 185 of 199 QUESTION # 094 Given the following:

A severe storm has passed through the area The following items were damaged and are reported out of service Electrical Feeder Mishicot -122 Electrical Feed East Krok-242 Siren Site P 002 Siren Site P 003 Siren Site P 013 Siren Site P 014 Which of the following is the Total Lost Coverage due to the storm? (See provided reference) 24.1%

33.4%

34.5%

38.5%

ANSWER:

B.

REFERENCE:

LI-AA-102-1001, Regulatory Reporting Rev 5 Proposed reference to be provided to the applicants during examination:

LI-AA-102-1001, Regulatory Reporting, Attachment 5, Pages 15 and 16 (69 and 70 of 97)

NEW HIGHER K/A: G2.1.2 Conduct of Operations: Knowledge of operator responsibilities during all modes of plant operation.

LEARNING OBJECTIVE:

Learning Objectives were derived from the tasks contained in NUREG-1122 Rev 2, Knowledge and Abilities Catalog for Nuclear Power Plant Operator: Pressurized Water Reactors. (2.1.2)

EXPLANATION:

% Population Feeder Siren Site# Covered Mishicot-122 P 003 1.1 Mishicot-122 P 005 1.4 Mishicot-122 P 008 3.4 Mishicot-122 P 009 2.0 Mishicot-122 P 010 1.7 Total 9.6 East Krok-242 K 003 2.1 East Krok-242 K 004 1.5 East Krok-242 K 007 1.0 Tuesday, December 23, 2014 185

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 186 of 199 Total 4.6 P 002 4.8 P 013 3.4 P 014 1.7 P-MultiFail P013 & P014 14.4 INCORRECT: See above. Adding (Mishicot-122) + (East Krok-242) + (P 002) + (P013) + (P014) = An incorrect answer: (9.6) + (4.6) + (4.8) +

(3.4) + (1.7) = 24.1. Plausible if the student misinterprets the note on how to deal with the P-MultiFail sirens.

CORRECT: See above. Adding (Mishicot-122) + (East Krok-242) + (P 002) + (P-MultiFail P013 & P014) = The correct answer: (9.6) + (4.6) +

(4.8) + (14.4) = 33.4 INCORRECT: See above. Adding (Mishicot-122) + (East Krok-242) + (P 002) + (P 003) + (P-MultiFail P013 & P014) = An incorrect answer: (9.6)

+ (4.6) + (4.8) + (1.1) + (14.4) = 34.5. Plausible if the student doesn't note that siren site P 003 is also part of the Mishicot-122 electrical feeder and counts it twice.

INCORRECT: See above. Adding (Mishicot-122) + (East Krok-242) + (P 002) + (P013) + (P014) + (P-MultiFail P013 & P014) = An incorrect answer: (9.6) + (4.6) + (4.8) + (3.4) + (1.7) + (14.4) = 38.5. Plausible if the student misinterprets the note on how to deal with the P-MultiFail sirens.

Tuesday, December 23, 2014 186

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 187 of 199 QUESTION # 095 Given the following:

Unit 2 is being refueled following a complete core offload.

Any deviation from the specified core refueling approved sequence, while transporting fuel to or from the Spent Fuel Pool or the core, requires the concurrence of the before any changes are made?

Reactor Engineering Duty and Call Supervisor Core Loading Supervisor and the Shift Manager Assigned Reactor Engineer and the Shift Manager Assigned Reactor Engineer and the Core Load Supervisor ANSWER:

D.

REFERENCE:

RP 1C, Refueling, Section 3.2.2 Rev 74 BANK FUNDAMENTAL K/A: G2.1.35 Conduct of Operations: Knowledge of the fuel-handling responsibilities of SROs.

LEARNING OBJECTIVE:

DESCRIBE the procedures which govern Fuel Handling Operations.

Description should include significant prerequisites, precautions, and notes associated with each operating procedure requiring consideration by Licensed and Non-Licensed Operators. (112.01.LP0285.002)

EXPLANATION:

INCORRECT: This is the position notified of issues during the fueling sequence. Plausible as this person is notified when issues arise.

INCORRECT: The first person is correct, but the Shift Manager is only required to be notified of the change. Plausible because the change is required to go through the Shift Manager for notification purposes.

INCORRECT: The first person is correct, but the Shift Manager is only required to be notified of the change. Plausible because the change is required to go through the Shift Manager for notification purposes.

CORRECT: This is the correct answer per section 3.2.2.

Tuesday, December 23, 2014 187

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 188 of 199 QUESTION # 096 Given the following:

P-32B, Service Water Pump, is isolated with a Clearance Order for preventative maintenance The Clearance Order for P-32B has been released and work has begun Mechanical Maintenance has contacted the Work Control Center with a change in the scope of work for P-32B Maintenance Planning has generated a new Work Order for the additional work What action(s) will be needed to add the new Work Order to the existing Clearance Order for P-32B per OP-AA-101-1000, Fleet Clearance and Tagging?

Only an SRO, with a peer check, is required to verify the current clearance isolations support the new work and approve the Clearance Order.

An SRO or Clearance Owner, with a peer check, will verify the current Clearance isolations support the new work and approve the Clearance Order.

A Clearance Owner and Lead Mechanic from the cognizant work group are required to approve the Work Order is supported by the current Clearance Order.

An SRO, with a peer check, and a Clearance Owner from the cognizant work group will verify the Work Order is supported by the current Clearance isolations.

ANSWER:

D.

REFERENCE:

OP-AA-101-1000, Fleet Clearance and Tagging, Step 4.9.5.A.(1) and B.(1) Rev 10 NEW FUNDAMENTAL K/A: G2.2.19 Equipment Control: Knowledge of maintenance work order requirements.

LEARNING OBJECTIVE:

Knowledge of maintenance work order requirements. (2.02.19)

EXPLANATION:

INCORRECT: See below. Plausible if the student does not understand the clearance order process and requirements.

INCORRECT: See below. Plausible if the student does not understand the clearance order process and requirements.

INCORRECT: See below. Plausible if the student does not understand the clearance order process and requirements.

CORRECT: Per OP-AA-101-1000 Fleet Clearance and Tagging step 4.9.5 an SRO with a peer checker who is also a reviewer must agree that Tuesday, December 23, 2014 188

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 189 of 199 the clearance isolation points meet the change in scope. The work group Clearance owner also signs as the second person on the work order the say the clearance is good for the scope of work.

Tuesday, December 23, 2014 189

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 190 of 199 QUESTION # 097 Given the following:

The following radiological conditions exist for an area in the plant:

General dose rates range from 25-45 mrem/hr Measurements taken on pipes and valves are:

Point 1 is 100 mrem/hr at 30 cm Point 2 is 500 mrem/hr at 30 cm Point 3 is 1100 mrem/hr at 30 cm A Non-Licensed Operator (NLO) needs to enter this area to implement Emergency Operating Procedure steps Based on the above information which of the following states:

(1) what should be the radiological posting for this room AND (2) who can authorize entry by the NLO?

(1) Very High Radiation Area (2) Only the RP Manager (1) Very High Radiation Area (2) The Plant General Manager and Shift Manager (1) Locked High Radiation Area (2) The Shift Manager if no RP supervisor can be reached (1) Locked High Radiation Area (2) Only the Plant General Manager if no RP supervisor can be reached ANSWER:

C.

REFERENCE:

RP-AA-103-1002, High Radiation Area Controls Rev 2 MODIFIED: 2010 Beaver Valley NRC Exam HIGHER K/A: G2.3.13 Radiation Control: Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

LEARNING OBJECTIVE:

Learning Objectives were derived from the tasks contained in NUREG-1122 Rev 2, Knowledge and Abilities Catalog for Nuclear Power Plant Operator: Pressurized Water Reactors. EXPLANATION:

INCORRECT: Per RP-AA-103-1002, Very High Radiation Areas are defined as an area, accessible to individuals, where radiation levels could result in an individual receiving an absorbed dose in excess of 500 rads in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 1 meter from a radiation source or from any surface that the radiation penetrates. The RP Manager and Plant General Manager permission is required to enter a Very High Radiation Area. Plausible because the candidate could confuse 500 R/hr with 500 mrem/hour.

Tuesday, December 23, 2014 190

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 191 of 199 INCORRECT: Per RP-AA-103-1002, Very High Radiation Areas are defined as an area, accessible to individuals, where radiation levels could result in an individual receiving an absorbed dose in excess of 500 rads in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 1 meter from a radiation source or from any surface that the radiation penetrates. The RP Manager and Plant General Manager permission is required to enter a Very High Radiation Area. Plausible because the candidate could confuse 500 R/hr with 500 mrem/hour, and cannot recall whose permission is required for entry.

CORRECT: Per RP-AA-103-1002, Locked High Radiation Areas are defined as an area, accessible to individuals where radiation level could result in an individual receiving a deep dose equivalent in excess of 1,000 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 centimeters from the radiation source or from any surface that the radiation penetrates. During an emergency the SM can provide authorization to make the entry and notify RP supervisor as soon as practical. The required key(s) to gain access can be issued from the control room by the SM.

INCORRECT: Per RP-AA-103-1002, Locked High Radiation Areas are defined as an area, accessible to individuals where radiation level could result in an individual receiving a deep dose equivalent in excess of 1,000 mrem in 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> at 30 centimeters from the radiation source or from any surface that the radiation penetrates. The second half is incorrect, During an emergency the SM can provide authorization to make the entry and notify RP supervisor as soon as practical. The required key(s) to gain access can be issued from the control room by the SM. Plausible if the student cannot recall the entry requirements.

Tuesday, December 23, 2014 191

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 192 of 199 Original Question:

The following radiological conditions exist in the letdown degassifier valve room:

general dose rates range from 25-45 mrem/hr measurements taken on pipes and valves are:

point 1 is 100 mrem/hr at 30 cm point 2 is 500 mrem/hr at 30 cm point 3 is 1100 mrem/hr at 30 cm the room needs to be accessed by an NSO no RWP has been authorized for this area no HP supervisor can be reached Based on these conditions what should be the radiological posting for this room, and who can authorize a RWP to allow entry by the NSO?

High radiation area, only a HP supervisor High radiation area, the station director and the shift manager Tech spec locked high radiation area, the shift manager if no HP supervisor can be reached Tech spec locked high radiation area, only the station director if no HP supervisor can be reached Tuesday, December 23, 2014 192

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 193 of 199 QUESTION # 098 Which of the following completes the statement?

(1) The Technical Specification 3.4.16 Reactor Coolant System - Specific Activity, limit for dose equivalent I-131 in the Reactor Coolant System is AND (2) During an emergency, in order to declare an ACTUAL LOSS of the Fuel Clad Barrier, the dose equivalent I-131 in the Reactor Coolant must be greater than as identified in EPIP-1.2, Emergency Classification, Fission Product Barriers Table.

(1)0.5 µCi/gm (2) 300 µCi/gm (1) 0.5 µCi/gm (2) 1500 µCi/gm (1) 50 µCi/gm (2) 300 µCi/gm (1) 50 µCi/gm (2) 1500 µCi/gm ANSWER:

A.

REFERENCE:

TS 3.4.16, Reactor Coolant System (RCS), RCS Specific Activity rev 1-233 EPIP 1.2, Emergency Classification Rev 52 MODIFIED: 2010 Turkey Point NRC Exam FUNDAMENTAL K/A: G2.3.14 Radiation Control: Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

LEARNING OBJECTIVE:

Learning Objectives were derived from the tasks contained in NUREG-1122 Rev 2, Knowledge and Abilities Catalog for Nuclear Power Plant Operator: Pressurized Water Reactors. (2.3.14)

EXPLANATION:

CORRECT: This is the value called out in Technical Specifications, as well as the primary coolant activity level which defines a loss of the fuel clad barrier.

INCORRECT: The value called out in Technical Specifications is correct.

The value which defines a loss of the fuel clad barrier is not correct.

Plausible because the value is for defining a loss of the fuel clad barrier based on Xe-133.

INCORRECT: The value for 3.4.16 is incorrect. The primary coolant activity level which defines a loss of the fuel clad barrier is correct.

Plausible because 50 µCi/gm is the Technical Specification limit for iodine spiking.

Tuesday, December 23, 2014 193

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 194 of 199 INCORRECT: The value for 3.4.16 is incorrect. The value which defines a loss of the fuel clad barrier is not correct.. Plausible because 50

µCi/gm is the Technical Specification limit for iodine spiking and the value is for defining a loss of the fuel clad barrier based on Xe-133.

Tuesday, December 23, 2014 194

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 195 of 199 Original Question:

Which ONE of the following completes the following statements?

The Technical Specification 3.4.8, Reactor Coolant System - Specific Activity, limit for dose equivalent I-131 in the Reactor Coolant System is During an emergency, in order to declare an ACTUAL LOSS of the fuel clad barrier, the dose equivalent I-131 in the Reactor Coolant must be at least in accordance with 0-EPIP-20101, Duties of Emergency Coordinator - Fission Product Barrier Table Worksheet 1 µCi/gm; 60 µCi/gm 1 µCi/gm; 300 µCi/gm 100/E-bar; 60 µCi/gm7 100/E-bar; 300 µCi/gm Tuesday, December 23, 2014 195

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 196 of 199 QUESTION # 099 Given the following:

Unit 2 is experiencing an event which has resulted in an ALERT EAL being exceeded and recognized by the crew After five (5) minutes of actions being taken by the crew, the ALERT EAL is no longer met and conditions for an UNUSUAL EVENT now are present No emergency declaration has been made Which of the following identifies the actions required by the Emergency Director?

Declare an ALERT, terminate the ALERT, declare an UNUSUAL EVENT Declare an ALERT, then de-escalate the Emergency Classification to an UNUSUAL EVENT Declare an UNUSUAL EVENT, and make a non-emergency update within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> report to document the ALERT Declare an UNUSUAL EVENT, and ensure the ALERT which was present is communicated to the NRC, via the Event Notification Worksheet ANSWER:

B.

REFERENCE:

EPIP 1.2, Emergency Classification Rev 52 BANK: 2011 Salem NRC Exam FUNDAMENTAL K/A: G2.4.38 Emergency Procedures / Plan: Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

LEARNING OBJECTIVE:

Learning Objectives were derived from the tasks contained in NUREG-1122 Rev 2, Knowledge and Abilities Catalog for Nuclear Power Plant Operator: Pressurized Water Reactors.

EXPLANATION:

INCORRECT: This is the correct first step. The second step is incorrect, since there is no termination in the event. The third step would be correct only when the event no longer meets the level of an EAL and then once again met the level at a later time, which is not the case in this scenario. Plausible if the student has a misconception of de-escalation versus termination of an event.

CORRECT: This is the correct answer, because the original event is declared, the event never terminated, therefore, a de-escalation is required to the UNSUAL EVENT classification.

INCORRECT: The declaration of an UNUSUAL EVENT would be incorrect and the non-emergency notification is also incorrect. Plausible if the student has a misconception of a missed classification.

Tuesday, December 23, 2014 196

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 197 of 199 INCORRECT: The declaration of an UNUSUAL EVENT would be incorrect and the ALERT communication is also incorrect. Plausible if the student has a misconception of a missed classification.

Tuesday, December 23, 2014 197

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 198 of 199 QUESTION # 100 In accordance with EPIP 1.1, Course of Actions, which of the following identifies an action that may be delegated by Emergency Director (ED)?

Approval of dose extensions Authorizing the use of potassium iodide Issuing Protective Action Recommendations Directing assembly, accountability and evacuation ANSWER:

D.

REFERENCE:

EPIP 1.1, Course of Action, Section 2.1.5 Rev 71 - Withheld from public disclosure FUNDAMENTAL BANK K/A: G2.4.40 Emergency Procedures / Plan: Knowledge of SRO responsibilities in emergency plan implementation.

LEARNING OBJECTIVE:

Learning Objectives were derived from the tasks contained in NUREG-1122 Rev 2, Knowledge and Abilities Catalog for Nuclear Power Plant Operator: Pressurized Water Reactors.

EXPLANATION:

INCORRECT: The approval of dose extensions is not an activity that can be delegated. Plausible if the student has a misconception of actions/duties which may be delegated by the ED.

INCORRECT: The authorization of potassium iodide use is not an activity that can be delegated. Plausible if the student has a misconception of actions/duties which may be delegated by the ED.

INCORRECT: The issuing of protective action recommendations not an activity that can be delegated. Plausible if the student has a misconception of actions/duties which may be delegated by the ED.

CORRECT: This is an action/duty that the ED can delegate. This is normally an Ops Coordinator job (Attachment A, Command and Control Turnover Sheet)

Tuesday, December 23, 2014 198

2015 Point Beach Nuclear Plant Initial License Exam Answer Key Page 199 of 199 001 D 021 D 041 B 061 B 081 D 002 D 022 D 042 D 062 A 082 B 003 C 023 B 043 A 063 B 083 A 004 D 024 A 044 C 064 D 084 D 005 D 025 B 045 D 065 A 085 B 006 C 026 B 046 B 066 A 086 C 007 D 027 D 047 A 067 C 087 D 008 B 028 B 048 B 068 C 088 A 009 D 029 C 049 D 069 B 089 B 010 C 030 B 050 D 070 B 090 B 011 B 031 C 051 B 071 D 091 C 012 C 032 D 052 D 072 C 092 D 013 B 033 C 053 C 073 A 093 C 014 C 034 A 054 D 074 C 094 B 015 A 035 B 055 C 075 A 095 D 016 C 036 C 56 B 076 C 096 D 017 B 037 C 057 D 077 B 097 C 018 B 038 D 058 B 078 D 098 A 019 D 039 B 059 C 079 B 099 B 020 A 040 B 060 D 080 C 100 D Tuesday, December 23, 2014 3:13:48 PM 1