Text

2015 Fermi Power Plant Initial License Examination Proposed Written Exam
	ML15287A053
Person / Time
Site:	Fermi
Issue date:	10/13/2015
From:	; Division of Reactor Safety III
To:	; Detroit Edison, Co
	Zoia, C D
Shared Package
ML15267A808	List: ML15282A468; ML15286A470; ML15286A472; ML15287A043; ML15287A053;
References
	Download: ML15287A053 (218)
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ID: R01 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW The plant is operating with the following indications:

B21-R612 Jet Pump Total Flow Recorder 87.4 Mlbs/hr C51-R608A Rx Power APRM 1 100%B31-R621A/B RR MG SET SPEED CONTROLLERs 73% (controlling in AUTO)

Based on the above starting conditions and assu ming no operator action in response to the transient, opening the field breaker for 'A' Reac tor Recirc Pump will cause Reactor Power to lower. The power reduction is due to _______.A.'A' Reactor Recirc pump coasting down ONLY.

B.'A' Reactor Recirc Pump coasting down PLUS a speed reduction in 'B' Reactor Recirc pump because of #4 limiter.C.'A' Reactor Recirc Pump coasting dow n PLUS a speed reduction in 'B' Reactor Recirc pump because of #2 and #3 limiters.D.'A' Reactor Recirc Pump coasting dow n PLUS a speed reduction in 'B' Reactor Recirc pump because of limiter enforcem ent, placing the core in the SCRAM region.Answer:A ILO 2015 Written Page: 1 of 218 25 June 2015 Answer Explanation:

Limiter 4 is the only runback that would actuate.

At a setting of 75%, it would have no effect on the reduction on flow. Therefore the coast down of A pump is the only thing lowering Rx power.

Distractor Explanation:B.This answer is incorrect because the setpoint of Limiter #4 is 75%, this answer is plausible if the examinee incorrectly thought #4 lim iter as a lower setpoint.C.This answer is incorrect and because Limiters #2 and

3 are cut out. This answer is plausible if the examinee incorrectly believes #2 & #3 limiters will enforce. D.This answer is incorrect because Limiter #2 and #3 are cut out, this answer is plausible if the examinee incorrectly believes flow would lower due to limiter enforcement.

Reference Information:

23.138.01 Section 1.0 Description (Limiter

2,#3,#4 on page 8) - explains limiters Plant Procedures 23.138.01 NUREG 1123 KA Catalog Rev. 2295001 AK3.Knowledge of the reasons for the followi ng responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION :295001 AK3.02Reactor power response 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam Early Review Questions for NRC 2015 ILO ILO 2015 Written Page: 2 of 218 25 June 2015 ID: R02 Points: 1.00Difficulty:

0.00 Level of Knowledge:

Low Source: NEW Following a loss of 120 kV and a reactor SCRAM, the following indications exist on panel H11-P809: Breaker Positions BUS 64B POS B8 CLOSED BUS 64B POS B6 TRIPPED BUS 11EA POS EA3 CLOSED BUS 12EB POS EB3 OPEN BUS 64C POS C8 CLOSED BUS 64C POS C6 TRIPPED Based on these indications (1) what event has occurred, and (2) what actions should the operating crew take in addition to entering AOP 20.300.120kV?A.(1) Bus lockout on BUS 64B (2) Perform 20.300.72B, Loss Of Bus 72B ONLY.B.(1) Bus lockout on BUS 64C (2) Perform 20.300.72C, Loss Of Bus 72C ONLY.C.(1) EDG 12 failure due to a fault with EDG 12 (2) Perform 20.307.01, Emergency Diesel Generator Failure AND 20.300.72C, Loss Of Bus 72C.D.(1) EDG 11 failure due to a fault with EDG 11 (2) Perform 20.307.01, Emergency Diesel Generator Failure AND 20.300.72B, Loss Of Bus 72B.Answer:C ILO 2015 Written Page: 3 of 218 25 June 2015 Answer Explanation:

20.300.120kV directs 20.307.01 and 20.300.72C if the EDG 12 output breaker fails to close (Condition K). Additionally, no bus lockout is indicated so 20.

307.01 would be effective if the cause of the EDG failure is correctable.

Distracter Explanation:A.Is incorrect and plausible because there is no i ndication of a 64B BUS LOCKOUT (indicated by BUS isolation). If there was a 64B BUS LOCK OUT, entering 20.300.72B would be correct.B.Is incorrect and plausible because there there is no indication of a 64C BUS LOCKOUT (indicated by BUS isolation). If there was a 64C BUS LO CKOUT, entering 20.300.72C would be correct.D.Is incorrect and plausible because the BUS 11EA POS EA3 is CLOSED so EDG 11 has started and loaded. If EDG 11 had failed then 20.307.

01 and 20.300.72B would be correct.

Reference Information:

20.300.120kv, page 9, Condition K.

Plant Procedures 20.300.120kv NUREG 1123 KA Catalog Rev. 2295003 AA1. Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER :

295003 AA1.02 Emergency generators 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 4 of 218 25 June 2015 ID: R03 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW-CONTRACT At 1030, a loss of offsite power occurred. AOP 20.300.Offsite, Loss of Offsite Power, is being implemented to mitigate the electrical system event.

At 1145 (current time), electrical system indications and conditions are as follows: *All EDGs are running loaded. *120kV and 345kV switchyard components are damaged and are being evaluated by personnel for restoration. *10D72 BOP 260/130V BATTERY 2PC TROUBLE is in alarm.

9D17 DIV I ESS 130V BATTERY 2PA TROUBLE alarmed but is now clear. *10D68 DIV II ESS 130V BATTERY 2PB TROUBLE alarmed but is now clear.

If the electrical system is operating within its design bases, based on the indications, (1) what is the status of the station batteri es and (2) what is their ability to supply loads during the event?A.(1)Battery 2PC is below minimu m cell voltage, loads are de-energized.(2)Batteries 2PA and 2PB are above minimum cell voltage, loads are supplied adequate power to operate.B.(1)Battery 2PC is above minimum cell voltage, BOP loads are supplied adequate power to operate.(2)Batteries 2PA and 2PB are above minimum cell voltage, ESF loads are supplied adequate power to operate.C.(1)Battery 2PC is above minimum cell voltage, BOP loads should be stripped in order to supply adequate power to remaining loads.(2)Batteries 2PA and 2PB are above minimum cell voltage, ESF loads are supplied adequate power to operate.D.(1)Battery 2PC is below minimum cell voltage, BOP loads should be stripped in order to supply adequate power to remaining loads.(2)Batteries 2PA and 2PB are below mi nimum cell voltage, ESF loads should be stripped in order to supply adequate power to remaining loads.Answer:C ILO 2015 Written Page: 5 of 218 25 June 2015 Answer Explanation:

Based on the design of the BOP batteries, under loaded conditions the batteries can supply loads for up to 90 minutes before falling below minimum cell vo ltage with no charger available. Based on the conditions of the 120kv switchyard, currently no BO P charger is available. The procedure requires stripping the DC BOP busses after one hour if the DC BOP chargers remain unavailable. This is to retain the batteries available for DC control power use up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ESF batteries are fully available with chargers energized due to the EDG power being availabl

e. Alarms indicate battery chargers are in service for the ESF batteries as directed by the procedure.

Distracter Explanation:A.Is incorrect and plausible because the examinee c ould incorrectly interpret the BOP battery trouble alarm as a loss of the battery voltage. The volt age alarm is above the minimum cell voltage. This would also be supported by the indication that o ffsite power cannot be restored. ESF chargers are energized as indicated by the alarm status whic h also enhances the plausibility of the distracter. B.Is incorrect and plausible because the examinee could incorrectly interpret that since EDG power is available all available battery chargers bot h ESF and BOP are energized. ESF chargers are energized as indicated by the alarm status whic h also enhances the plausibility of the distracter. D.Is incorrect and plausible because the examinee could incorrectly interpret the indication and application of the knowledge of the design basis of

1.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months s

to the BOP battery and conclude that since one hour has passed without stripping the BO P battery, that the cells are below minimum voltage.Reference Information:

R32-00 DC ELECTRICAL SYSTEM DBD pages 26-27 Criteria for BOP batteries Plant Procedures 20.300.Offsite R32-00 DESIGN BASIS DOCUMEN T FOR DC ELECTRICAL SYSTEM NUREG 1123 KA Catalog Rev. 2295004 AA2.Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER :295004 AA2.03Battery voltage 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam

Early Review Questions for NRC 2015 ILO ILO 2015 Written Page: 6 of 218 25 June 2015 ID: R04 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: BANK:802-2003-0006-008 With the plant operating at 90% power, a F eedwater Controller Failure - Maximum Demand occurs. Automatic protection from this transient is assured by a scram generated as a result of

________.A.OPRM - Upscale direct scram, whic h prevents exceeding the MCPR Safety LimitB.Reactor Water Level High - Level 8 Main Turbine Trip indirect scram, which prevents exceeding the MCPR Safety LimitC.Reactor Vessel Steam Dome Pressure - High direct scram, which prevents exceeding the Reactor Coolant System Pressure Safety LimitD.APRM Simulated Thermal Power Upscale direct scram, which prevents exceeding the Reactor Coolant System Pressure Safety LimitAnswer:B ILO 2015 Written Page: 7 of 218 25 June 2015 Answer Explanation:

The feedwater and main turbine high water level tr ip instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event. The Level 8 trip indirectly initiate s a reactor scram (above 30% RTP) from the main turbine trip and trips the feedwater pumps, thereby te rminating the event. The reactor scram mitigates the reduction in MCPR.

Distracter Explanation:A.This is plausible and incorrect because this prot ects against thermal hydr aulic power oscillations.C.This is plausible and incorrect because this prot ects against any event that result in a reactor pressure increase.D.This is plausible and incorrect because this prov ides protection against a loss of feedwater heating and ensures the MCPR SL is not exceeded.

Reference Information:T.S.B. 3.3.2.2 (pg 3.3.2.2-2) APPLICABLE SAFETY ANALYSES NUREG 1123 KA Catalog Rev. 2G2.2.38Knowledge of conditions and lim itations in the facility license.295005 AK1Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP:295005 AK1.02 Core thermal limit considerations.

Technical Specifications3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 8 of 218 25 June 2015 ID: R05 Points: 1.00Difficulty:

3.50 Level of Knowledge:

High Source: BANK: 315-0048-A018-001 The plant was operating at 15% power during a startup. Due to an electrical malfunction, ALL Turbine Control Valves opened resulting in a Reactor Scram. No operator actions have been taken.Which one of the following lists the systems av ailable for Decay Heat Removal immediately following the scram?A.Safety Relief Valves and Reactor Water Cleanup ONLY.

B.Turbine Bypass Valves, Safety Relief Valves and Reactor Water Cleanup ONLY.C.Main Steam Line Drain Valves, Turbine Bypass Valves and Safety Relief Valves ONLY.D.Main Steam Line Drain Valves, Turbine Bypass Valves, Safety Relief Valves and Reactor Water Cleanup.Answer:A ILO 2015 Written Page: 9 of 218 25 June 2015 Answer Explanation:

With power at 15%, the Mode Switch is in RUN.

TCVs opening lowers RPV Pressure. When pressure reaches 756 psig with Mode Switch in RUN, a GP1 PCIS isolation occurs and a Scram. With MSIVs closed, only SRVs and RWCU are immediately available to remove decay heat.

Distracter Explanation:B. Is incorrect and plausible because Tu rbine Bypass Valves are not available.C.Is incorrect and plausible because Turbine By pass Valves and MSL Drains are not available.D.Is incorrect and plausible because Turbine By pass Valves and MSL Drains are not available.

Reference Information:

23.601 Main Steam Line Pressure - Low 23.601 Main Steam Line Isolation Valve - Closure Limit Switches

20.000.21 Bases Mode Switch interlocks Plant Procedures 20.000.21 20.000.21 Bases NUREG 1123 KA Catalog Rev. 2295006 AK1.Knowledge of the operational implications of the following concepts as they apply to SCRAM :295006 AK1.01Decay heat generation and removal 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 10 of 218 25 June 2015 ID: R06 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW-CONTRACTThe plant was operating at 100% power when a fire occurred in the Owner Controlled Area. The fire engulfed several buildings in cluding the onsite vehicle garage.

The following timeline of events subsequently occurred:10:30Heavy black smoke is drawn into the ventilation intake making the Control Room uninhabitable.11:15All offsite power is lost.

Based on the above plant conditions, at 10:30 cont rol of the plant would have shifted to the

__________________.A.Remote Shutdown Panel and remained there for the duration of the eventB.Dedicated Shutdown Panel and remained there for the duration of the eventC.Remote Shutdown Panel and subsequently to the Dedicated Shutdown Panel at 11:15D.Dedicated Shutdown Panel and subsequent ly to the Remote Shutdown Panel at 11:15Answer:C ILO 2015 Written Page: 11 of 218 25 June 2015 Answer Explanation:

Initially at 10:30 the correct location for remote cont rol of the plant due to an uncomplicated control room evacuation is the Remote Shutdown Panel. At time 11:15 control of the plant is required to shift to the Dedicated Shutdown Panel since the Remote Shutdown System is designed for an uncomplicated control room evacuation event and as such does not serve to mitigate the full range of design basis events.

Therefore, transfer to the Dedicated Shutdown Panel is required since this system provides a means of restoring AC power to the Div 1 ESF and BOP busse s as well as the necessary system controls and instrumentation to maintain the reactor core co vered in the event of a Control Room evacuation coincident with a loss of offsite power.

Distracter Explanation:A.Is incorrect and plausible because the examinee coul d incorrectly assume that the Remote Shutdown Panel was designed for this type of event. The candi date could fail to recognize that the Remote Shutdown System is designed for a control room ev acuation event (short term duration) and as such does not serve to mitigate the full range of design bas is events for which the Main Control Room was designed. Furthermore the candidate could also fa il to recognize that the Dedicated Shutdown System provides a means of restoring AC power to the Div 1 ESF and BOP busses. The panel also provides necessary system controls and instrumentation to maintain the reactor core covered. Power restoration may become necessary following a fire or loss of offsite power.B.Is incorrect and plausible if the examinee incorrect ly assumed that control of the plant from the Dedicated Shutdown Panel was required for all Main Control Room evacuation events that are a result of an onsite fire. The candidate could fail to recognize that control fr om the Dedicated Shutdown Panel is only required for fires in a 3L "zone of conc ern" that has the potential to impact safe plant shutdown capability due to multiple hot shorts and/or open circuits causing loss of control from the control room. D.Is incorrect and plausible if the examinee incorrect ly assumed that control of the plant from the Dedicated Shutdown Panel was required for all Main Control Room evacuation events that are a result of an onsite fire. The candidate could t hen conclude that, upon Loss of Offsite Power, the correct response would be to shift control to the Remote Shutdown Panel. The candidate could fail to recognize that the Dedicated Shut down System is designed to allow for remote control of the plant concurrent with a loss of offsite power event.

Reference Information:

20.000.19 (pg 2) Overide 1 20.000.19 BASES (pg 2) Overide 1 Bases

20.000.18 BASES (pg 2) - BASES

SUMMARY

ILO 2015 Written Page: 12 of 218 25 June 2015 Plant Procedures 20.000.18 20.000.19 20.000.18 Bases

20.000.19 Bases NUREG 1123 KA Catalog Rev. 2295016 AK2.Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following:295016 AK2.02Local control stations: Plant-Specific 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 13 of 218 25 June 2015 ID: R07 Points: 1.00Difficulty:

2.00 Level of Knowledge:

High Source: NEW-CONTRACT The plant is operating at 100% power with the following conditions: *#2 & #5 GSW Pumps are running. *#3 GSW Pump is OOS for motor replacement.

1. 4 GSW Pump has just tripped.

1. 6 GSW Pump is OOS for discharge strainer leak repair.

GSW Header Pressure steady at 70 psig with the P4100-F841, GSW Bypass Line Pressure Ctrl Vlv, 50% open. *RBCCW Heat Exchanger outlet temperature is 90°F and rising.

Which ONE of the following actions is requi red to stabilize the above conditions?A.Dispatch an operator to throttl e closed P4100-F841, GSW Bypass Line Pressure Ctrl Vlv, in order to raise GSW header pressure and restore adequate cooling water flow.B.Dispatch an operator to throttle open P4100-F840, GSW Flow Test Pressure Ctrl Vlv, in order to raise GSW header pressure and restore adequate cooling

water flow.C.Scram the reactor, trip the main turb ine and initiate Div 1 and 2 EECW in order to establish cooling to safety related equipment and restore GSW header

pressure to normal band.D.Increase cooling water flow using P 42-F400, RBCCW Temp Control Vlv in AUTO, or MANUAL if necessary, in or der to restore RBCCW Heat Exchanger outlet temperature and GSW header pressure in band.Answer:A ILO 2015 Written Page: 14 of 218 25 June 2015 Answer Explanation:

P4100-F841, GSW Bypass Line Pressure Ctrl Vlv is a dum p, or backpressure, control valve. Prior to the transient and trip of #4 GSW Pump, this valve was required to be open to maintain GSW header pressure due to 3 pumps being in excess of the plant cooling water requirements. With only 2 GSW pumps running, closing this valve will raise GSW header pre ssure and increase cooling water flow to the CCW heat exchangers cooled by GSW thereby lowering CCW cooling temperatures, such as the elevated

RBCCW Heat Exchanger outlet.

Distracter Explanation:B.Is incorrect and plausible because the examinee could incorrectly determine that, since the P4100-F841, GSW Bypass Line Pressure Ctrl Vlv, is at its 50% open limit then pressure control would logically transfer to the P4100-F840, GSW Flow Test Pressure Ctrl Vlv. This would be correct for a high pressure condition. The examinee could also incorrectly determine that GSW pressure control valves are throttled OPEN to raise GSW header pressure, however, the valves are actually backpressure control valves and must be throttled CLOSED to raise GSW header pressure and increase cooling water flow to supported CCW heat exchangers.C.Is incorrect and plausible because the candidate could incorrectly determine that, since GSW header pressure is below 65#, the plant and main turbine should be tripped, and EECW initiated, as directed

by the loss of GSW AOP. However, the loss of GSW AOP override statement only requires a plant scram if GSW header pressure cannot be restored AND MAINTAINED above 65 psig. The examinee should determine that throttling capacity exists to restore GSW header pressure without the need for these drastic actions.D.Is incorrect and plausible because the candidate c ould incorrectly determine that throttling open the P42-F400, RBCCW Temp Control Vlv in AUTO, or MANUAL, would corre ct the elevated RBCCW Heat Exchanger outlet temperature condition.

Although this action is directed by ARP 2D120, RBCCW HX DISCH TEMPERATURE HIGH/LOW for a High Temperature condition, the action is accompanied by a conditional statement to ma ke the adjustments while monitoring GSW header pressure. With GSW header pressure alr eady degraded, this action would not be prudent.

Reference Information:

ARP 7D14 (pg 1) GSW pressure low actions

23.131 (pg 23-25) GSW pressure control ILO 2015 Written Page: 15 of 218 25 June 2015 Plant Procedures 23.131 07D14 23.127 02D120 NUREG 1123 KA Catalog Rev. 2295018 AK3.Knowledge of the reasons for the followi ng responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:295018 AK3.06Increasing cooling water flow to heat exchangers.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 16 of 218 25 June 2015 ID: R08 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW The plant is operating at 100% power with the following Station Air Compressor lineup:East Station Air Compressor Off Center Station Air Compressor Running West Station Air Compressor Auto Following a seismic event, Bus 72A has been de-energized, and both Control Air Compressors auto start due to lowering air pressure. At 2 ps i below the Control Air Compressors auto start setpoint, all air header pressures begin to recove

r. Assuming no operator action, what is the status of the following Station Air valves?P50-F401P50-F402P50-F440 STATION AIR

TO TB HDR ISO VLV STATION AIR

TO NIAS ISO VLV DIV 1 CONTROL

AIR ISO VLV A.B.

C.

D.Answer:B ILO 2015 Written Page: 17 of 218 25 June 2015 CLOSED CLOSED OPEN OPEN OPEN CLOSED CLOSED OPEN CLOSED OPEN OPEN CLOSED Answer Explanation:

Based on logic prints listed & AOP 20.129.01, Control Air Compressors start at 85 psi therefore system press has to go to below 85 psi and then, per the questi ons stem pressure recovers. Based on this the 401 will get a closed signal, and the 402 and 440 will not get close signals.

Distracter Explanation:

Distracter are plausible based on not understanding setpoints or power supplies/auto-starts of compressors A.Is incorrect because 401 will get close signal at 85#.

This answer is plausible if the examinee incorrectly assumes the system pressure did not go low enough to cause isolations.C.Is incorrect because 402 will not get close signal until 75# and air header only went to 83#. This answer is plausible if the examinee incorrectly assumes the system pressure did go low enough to cause isolations.D.Is incorrect because 402 & 440 will not get close signal until 75# and air header only went to 83#.

This answer is plausible if the examinee incorre ctly assumes the system pressure did go low enough to cause isolations.

Reference Information:

I-2450-02,04,05 are not shown here, however they ar e included in development folder, the AOP actions match the actuations.

20.129.01 (pg 3) System actuation bas ed on pressure per action statement.

Plant Procedures 20.129.01 20.300.72A NUREG 1123 KA Catalog Rev. 2295019 AA1.Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:295019 AA1.04Service air isolations valves: Plant-Specific 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 18 of 218 25 June 2015 ID: R09 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: MODIFIED BANK: ID 26798 The plant is in HOT SHUTDOWN with the following conditions:*Reactor Pressure is 25 psig.*RPV Water Level is 250 inches and stable.

RHR Pump B is operating in Shutdown Cooling Mode.

When the following alarm occurs:*2D26, DIV II RHR SYSTEM LOW FLOW BYPASS INITIATED.

Without operator action which one of the followi ng describes (1) the effect on Reactor Water Level and (2) reason for that effect?A.(1) Reactor Water Level will remain the same.

(2) The RHR pumps min flow valve has opened; system flow rate is reduced.B.(1) Reactor Water Level will remain the same. (2) The RHR system flow is low, indicating a low heat load condition.C.(1) Reactor Water Level will rise. (2) The RHR system flow from the vessel is low, this indicates a flow imbalance.D.(1) Reactor Water Level will lower.

(2) The RHR pumps min flow valve has opened, rejecting Rx water to the Torus.Answer:D ILO 2015 Written Page: 19 of 218 25 June 2015 Answer Explanation:

The alarm 2D26 indicates the RHR pumps min flow valve has opened resulting in increased pump flow (two paths). The min flow valve will reject Rx wate r to the Torus causing level to lower until a low Rx water level isolation is reached.

OE: SOER 87-002, Inadvertent Draining of the Reactor Vessel to Suppression Pool at BWRs Distracter Explanation:A.This is plausible and incorrect because SDC is a closed loop system, however when the minflow valve is open it becomes a rejection path causing level to lower.B.This is plausible and incorrect because the examinee could identify the low flow alarm to be low flow in a closed loop system, which would have no effect on level.C.This is plausible and incorrect because the exam inee could identify the low flow alarm with RHR output from the vessel, and given a feed and bleed type system this could identify a condition that

causes level to rise.

Reference Information:

ARP 2D26 (pg 1) Action for this condition.

FOS M-5706-1 Green highlight of path for drain to torus (located in Development folder)

Plant Procedures 03D079 02D026IERs/ SOERs/ SERs/ SENs/ O&MR SOER 87-2 Inadvertant Draining of Reac tor Vessel to Suppression Pool at BWRs NUREG 1123 KA Catalog Rev. 2295021 AA2.Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING:295021 AA2.03Reactor water level 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 20 of 218 25 June 2015 ID: R10 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW-CONTRACT Core alterations are in progress with the plant in Mode 5. 'A' RHR pump is operating in the shutdown cooling mode and 'C' RHR pump is ava ilable in standby. 'B' and 'D' RHR pumps are out of service. The control room receives a report that spent fuel pool level is slowly lowering. A nuclear operator isolates the source of the leak and the level reduction is stopped. *2D1 Fuel Pool Level Low is alarmed.*2D13 Fuel Pool Cooling Trouble is alarmed.

FPCCU Pumps have tripped.

The control room determines that RPV level is 20.0 ft above the RPV flange using B21-R605, RPV Water Level Flood Up Range.

Which one of the following actions is required?A.Suspend movement of fuel assemblies within the RPV.B.Verify an alternate method of decay heat removal is available.

C.Verify two alternate methods of decay heat removal are available.

D.Initiate action to restore sec ondary containment to operable status.Answer:A ILO 2015 Written Page: 21 of 218 25 June 2015 Answer Explanation:

The examinee should correctly determi ne that Technical Specification 3.9.6 (Refueling Operations - RPV Water Level) requires suspension of movement of f uel assemblies with RPV level < 20 ft 6 inches. The completion time for the action is IMMEDIATELY.

Distracter Explanation:

B.Is incorrect and plausible because examinee may in correctly identify Techni cal Specification 3.9.7 actions require one verification that an alternate method of decay heat removal is available with a completion time of IMMEDIATELY if one RHR SDC subs ystem is not in operation. With level < 20 ft 6 inches, this LCO is not applicable.C.Is incorrect and plausible because the examinee c ould incorrectly identify Technical Specification

3.9.8 requires

verification of an alternate met hod of decay heat is available for each inoperable required RHR SDC subsystem. Although B and D are inoperable two still remain and even though RPV level is less than 20 ft 6 inches which makes the LCO applicable, the condition for entry is not currently met.D.Is incorrect and plausible because the examinee c ould incorrectly identify Technical Specification 3.9.7 and 3.9.8 have actions to IMMEDIATELY rest ore secondary containment to operable status if

the completion times of condition A in both LCOs are not met.

Reference Information:

T.S. 3.9.6 (pg 3.9-9) LCO and Immediate action.

NUREG 1123 KA Catalog Rev. 2G2.2.39Knowledge of less than one hour techni cal specification action statements for systems.295023Refueling Accidents Technical Specifications3.9.6 Reactor Pressure Vessel (RPV) Water Level 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 22 of 218 25 June 2015 ID: R11 Points: 1.00Difficulty:

3.00 Level of Knowledge:

Low Source: NEW-CONTRACT Which of the responses below best completes t he following regarding the operational implications of High Drywell Pressure?

In MODES 1, 2 and 3, the maximum allowable Dryw ell Pressure is __(1)__. This limit is based on __(2)__.A.(1) 2.25 psig (2) maintaining the resultant differential pressure below the maximum primary containment design differential pressure in the event of inadvertent drywell spray actuation.B.(1) 2.0 psig (2) maintaining the resultant differential pressure below the maximum primary

containment design differential pressure in the event of inadvertent drywell spray actuation.C.(1) 2.25 psig(2) maintaining the resultant peak primary containment accident pressure below

the primary containment design pressu re in the event of a Design Basis Accident (DBA).D.(1) 2.0 psig(2) maintaining the resultant peak primary containment accident pressure below

the primary containment design pressu re in the event of a Design Basis Accident (DBA).Answer:D ILO 2015 Written Page: 23 of 218 25 June 2015 Answer Explanation:

The examinee should correctly conclude that, in a ccordance with LCO 3.6.1.4, the maximum Drywell Pressure in the specified MODEs is 2.0 psig. The examinee should also conclude that the bases for this limit is to preserve the initial conditions assumed in the accident anal ysis for the Design Basis Accident (DBA) thus ensuring that the peak primary containment pressure does not exceed the design pressure of the primary containment pressure.

Distracter Explanation:A.Is incorrect and plausible because the examinee c ould fail to recognize the LCO 3.6.1.4 maximum allowable Drywell Pressure of 2.0 psig. Also, t he examinee could fail to recognize that the basis given in the second part of the distractor is for t he minimum allowable Drywell Pressure specified by LCO 3.6.1.4 in MODES 1, 2 and 3, which protects containment integrity by keeping external to internal drywell differential pressure below the maximum allowable design D/P.B.Is incorrect and plausible because the first part of the distractor does provide the correct maximum Drywell Pressure allowed by LCO 3.6.1.4 in the MO DEs specified. However, the examinee could fail to recognize that the basis given in the second par t of the distractor is for the minimum allowable Drywell Pressure specified by LCO 3.6.1.4 in MODES 1, 2 and 3, which protects containment integrity by keeping external to internal drywell differential pre ssure below the maximum allowable design D/P.C.Is incorrect and plausible because the examinee c ould fail to recognize the LCO 3.6.1.4 maximum allowable Drywell Pressure of 2.0 psig.

Reference Information:

T.S. 3.6.1.4 (pg 3.6-18) LCO statementT.S. 3.6.1.4 BASES (pg B3.3.1.4-1 to 2) Bases for limit.

Plant Procedures 03D081 NUREG 1123 KA Catalog Rev. 2295024 EK1.Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE:295024 EK1.01Drywell integrity: Plant-Specific Technical Specifications3.6.1.4Primary Containment Pressure 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.

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2.00 Level of Knowledge:

Low Source: NEW-CONTRACT The reactor is near the end of an operating cycl e with Power and Recirculation Flow at 100%.

If the plant responds as assumed in the Fermi 2 over-pressurization protection analysis, when at least two Average Power Range Monitors reach

________ , the reactor will scram in order to terminate a Main Steam Isolation Valve (MSIV) closure event.A.107.2% Neutron FluxB.113.5% Simulated Thermal Power C.118% Neutron Flux D.122.2% Simulated Thermal PowerAnswer:C ILO 2015 Written Page: 25 of 218 25 June 2015 Answer Explanation:

As the MSIV closure event is occurring, reactor pressure will be increasing. The increase in RPV pressure during reactor operation will compress the steam voids and result in a positive reactivity insertion, which then causes neutron flux and Thermal Power to increase. The Fermi 2 over-pressurization protection analysis conservatively assumes scram on the Average Power Range Monitor Neutron Flux-Upscale trip of 118% which, along with the SRVs, limits the peak RPV pressure to less than the ASME Code limits.

Distracter Explanation:A. Is incorrect and plausible if the candidate a ssumed that the 107.2% was the APRM setpoint. The 107.2% is the RBM HTSP at >82% reactor power which will only cause a Rod Withdrawl Block.B.Is incorrect and plausible because the examinee c ould fail to recognize that the APRM STP-Upscale trip provides protection against tr ansients where Thermal Power increases slowly and not for rapidly changing power events such as the MSIV closure and, therefore, assumes that since the APRM STP-Upscale trip is clamped at an upper limit of 113.

5%, which is lower than the APRM Neutron Flux-Upscale Function Allowable Value at 100% recirculat ion flow, that this function will terminate the MSIV closure event before the APRM Upscale Function.D. Is incorrect and plausible because the examinee c ould fail to recognize that the APRM STP-Upscale trip provides protection against tr ansients where Thermal Power increases slowly and not for rapidly changing power events such as the MSIV closure and then calculates the STP upscale trip for 100%

recirculation flow using the following formula: 0.62(W-W) + 60.2%.

Reference Information:

T.S.B. 3.3.1.1 (pg B 3.3.1.1-14 to 15) Bases for MSIC Closure Plant Procedures 03D093 03D073 NUREG 1123 KA Catalog Rev. 2295025 EK2.Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:295025 EK2.09Reactor power Technical Specifications3.3.1.1Reactor Protection Syst em (RPS) Instrumentation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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3.00 Level of Knowledge:

Low Source: NEW-CONTRACT Which of the responses below best completes the following statement regarding the critical parameter for the Heat Capacity Limit (HCL) curve and the basis behind why the reactor must be Emergency Depressurized when the curve is exceeded?

The reactor is Emergency Depressurized when

__(1)__ exceeds the HCL to ensure that the

__(2)__ is not exceeded before the depressu rization of the RPV is complete.A.(1) Drywell Temperature (2) Drywell Design TemperatureB.(1) Drywell Temperature (2) Pressure Suppression PressureC.(1) Torus Water Temperature (2) Torus Design TemperatureD.(1) Torus Water Temperature (2) Pressure Suppression PressureAnswer:C ILO 2015 Written Page: 27 of 218 25 June 2015 Answer Explanation:The critical parameter on the HCL curve is Torus Water Temperature. Emergency Depressurizing due to exceeding the HCL curve is based on not exceeding either the torus design temperature or the PCPL

before the depressurization of the RPV is complete.

Distracter Explanation:A.Is incorrect and plausible because the examinee c ould determine that the cr itical parameter on the HCL curve is Drywell Temperature and therefore conclude that Emergency Depressurizing due to exceeding the HCL curve is based on not exceeding the drywell design temperature before the

depressurization of the RPV is complete.B. Is incorrect and plausible because the examinee c ould determine that the cr itical parameter on the HCL curve is Drywell Temperature. The candidate could incorrectly conclude that Emergency

Depressurizing due to exceeding the HCL curv e is based on not exceeding the Pressure Suppression Pressure of containment rather than the PCPL limit before the depressurization of the RPV is complete.D.Is incorrect and plausible because the examinee could correctly recognize that the critical parameter on the HCL curve is Torus Water Temperature. However, the candidate could incorrectly conclude that Emergency Depressurizing due to exceeding the HCL curve is based on not exceeding the

Pressure Suppression Pressure of containm ent rather than the PCPL limit before the depressurization of the RPV is complete.

Reference Information:

BWROG EPGs/SAGs, Appendix B (pg B-7-24 & B-17-16 to 17) Bases of HCL & Emergency Depressurization.

Plant Procedures 29.100.01 SH 2 BWROG EPG App B NUREG 1123 KA Catalog Rev. 2295026 EK3.Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:295026 EK3.01Emergency/normal depressurization 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.

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0.00 Level of Knowledge:

High Source: NEW-CONTRACT At which of the following is the Drywell Tem peratures and RPV pressures may PRV Water level indication become unreliable due to boiling in the instrument run? (See attached 29.100.01 Sheet 6.)Drywell TemperatureRPV pressureA.250°F25 PSIGB.325°F50 PSIG C.350°F150 PSIG D.375°F250 PSIGAnswer:B ILO 2015 Written Page: 29 of 218 25 June 2015 Answer Explanation:

Using the RPV Saturation Temperature curve, at 50 ps ig in the Reactor, the Drywell Temperature listed at which the instrument may become reliable due to boiling in the run is aboove 300°F.

Distracter Explanation:

Distracters a incorrect and plausible based in plotti ng the data in the RPV Saturation Temperature curve.

Reference Information:

29.100.01 SH 6 not include here, a copy is in the handout folder.

Plant Procedures 29.100.01 SH 6 NUREG 1123 KA Catalog Rev. 2295028 EA2.Ability to determine and/or interpret t he following as they apply to HIGH DRYWELL TEMPERATURE:295028 EA2.02Reactor pressure 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

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High Source: NEWA plant event has damaged the Reactor Building. The Operating Shift is executing the EOPs.*The reactor has scrammed however 12 rods are stuck at 04*RWL is 176 inches.

Reactor Pressure is 900 psig.

The MSIVs are open.The STA has plotted three points on the PSP based on valid data from MCR indication.

Based on this data, the Operating shift should pr epare for (1) which of the following courses of action and (2) why?A.(1) Depressurize using Bypass Valves ignoring cooldown rates.

(2) Because suppression pool water level will be low enough that steam discharged from the drywell into t he suppression pool may not be condensed.B.(1) Depressurize using Bypass Valves ignoring cooldown rates.

(2) Because suppression pool water level will correspond to the Minimum

Pressure Suppression Primary Containment Water Level. Below this level, the pressure suppression capability of the pr imary containment may be insufficient to accommodate an RPV breach by core debris.C.(1) Emergency Depressurize (2) Because suppression pool water level will be low enough that steam

discharged from the drywell into t he suppression pool may not be condensed.D.(1) Emergency Depressurize (2) Because suppression pool water level will correspond to the Minimum

Pressure Suppression Primary Containment Water Level. Below this level, the pressure suppression capability of the pr imary containment may be insufficient to accommodate an RPV breach by core debris.

ILO 2015 Written Page: 31 of 218 25 June 2015 Answer:C ILO 2015 Written Page: 32 of 218 25 June 2015 Answer Explanation:

Suppression pool water level must be maintained above the elevation of the Mark I/II downcomer vent openings or least 2 feet above the top of the Mark III horizontal vents to ensure that steam discharged from the drywell into the suppression pool following a primary system break will be adequately condensed. (Results of the Bodega Ba y Mark I containment tests indicate 95% steam condensation may be achieved from a vertical downcomer vent that di scharges at a level six inches above the suppression pool surface.) If suppression pool water level cannot be maintained above the specified minimum value, steam may not be adequately condensed and primary cont ainment pressure could exceed allowable limits. Since the RPV may not be kept at pressure when pressure suppression capability is unavailable, Emergency RPV Depressurization is required.

[Left most line] is the suppression pool water level corresponding to the elevati on of the downcomer vent openings or two feet above the horizont al vents, as appropriate. If suppr ession pool water level is below this elevation, the RPV may not be kept in a pre ssurized state since steam discharged through the vents may not be condensed.

[right most line] is the suppression pool water le vel corresponding to the Maximum Pressure Suppression Primary Containment Water Level. Above this elev ation, the pressure suppression capability of the primary containment may be insufficient to accommodate an RPV breach by core debris.

Distracters Explanation:

"(1) Depressurize using Bypass Valves ignoring cool down rates." - This answer is plausible and incorrect.

The examinee would choose this answer because it is a valid method to reduce primary pressure and it would put all that energy of the the depressurization in the condenser. However this is not the method allowed by the EOPs for this condition.

"(2) Because suppression pool water level will corre spond to the Minimum Pressure Suppression Primary Containment Water Level. Below this level, t he pressure suppression capability of the primary containment may be insufficient to accommodate an RPV breach by core debris." - This answer is plausible and incorrect. The examinee would c hoose this answer based on remembering the upper limit for TWL. IE the Maximum Pressure Suppression Primary Containment Water Level.

Reference Information:

BWR EPG Appendix B (pg B-7-49) and (pg B-17-64)

Plant Procedures BWROG EPG App B NUREG 1123 KA Catalog Rev. 2G2.4.18 Knowledge of the specific bases for EOPs295030 Low Suppression Pool Water Level 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

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High Source: NEW-CONTRACT Which one of the following represents the mini mum RPV level where (1) adequate core cooling exists, and (2) the maximum expected clad tem perature if SBFW pumps are the only source injecting?A.(1) -25" (2) 1500°FB.(1) -48" (2) 1500°FC.(1) -25" (2) 1800°FD.(1) -48" (2) 1800°FAnswer:A ILO 2015 Written Page: 35 of 218 25 June 2015 Answer Explanation:

The Minimum Steam Cooling RPV Water Level (MSCRWL) is defined as the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F. At Fermi 2 it is calculated to be

-25.Distracter Explanation:

All distracters are incorrect and plausible if the examinee does not understand the MSCRWL requirements.

The Minimum Zero-Injection RPV Water Level is defi ned to be the lowest RPV water level at which the covered portion of the reactor core will generate suffici ent steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1800 o F. Adequate Core Cooling is ensured as long as Core Spray requirements (5725 gpm) are satisfied and RPV water level can be restored and maintained at or above the elevation of t he jet pump suctions (-48 in).

Reference Information:

BWROG EPG (pg B-17-58) MCSRWL

MSCRWL Design Calc (pg 1)

Plant Procedures BWROG EPG App B NUREG 1123 KA Catalog Rev. 2295031 EK1.Knowledge of the operational implications of the following concepts as they apply to REACTOR LOW WATER LEVEL:295031 EK1.01Adequate core cooling NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 36 of 218 25 June 2015 ID: R17 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW-CONTRACT The plant has experienced a transient that result ed in a reactor scram with reactor power above 5%. The crew is using 29.ESP.03 for alternate control rod insertion.Which ONE of the following alternate control rod insertion methods in 29.ESP.03 could be made more effective by CLOSING C1100-F034, CRD Charging Water Header Isolation Valve?A.Increase CRD Cooling Water Differential PressureB.Scram Reset And Manual Scram Reinitiation C.Vent CRD Over Piston Volumes D.Manual Control Rod InsertionAnswer:D ILO 2015 Written Page: 37 of 218 25 June 2015 Answer Explanation:

Closing C1100-F034 is an effective means of raising dr ive water D/P. This is an option given in Section 3.0, Manual Control Rod Insertion, for the operator to use if unable to maintain sufficient drive water D/P with C1152-F003, CRD Drive/Clg Water PCV, and C11-K612, CRD Flow Controller.

Distracter Explanation:A.Is incorrect and plausible because closing C 1100-F034 would be an effective means of raising cooling water differential pressure, but if a scram signal is present and the CRD scram valves open, having C1100-F034 open would actually increase pre ssure on the underside of the drive piston the same as increasing cooling water pressure. Addi tionally, no procedural guidance is given for closing C1100-F034 in this section for that reason.B.Is incorrect and plausible because the examinee could incorrectly interpret the conditions under which operation of C1100-F034 is directed to be operat ed for this method and fail to recognize that, if reactor pressure is not available to scram control rods, then OPENING C1100-F034 would be effective in providing the under piston pressure nec essary for control rod insertion when the scram signal is re-initiated. C. Is incorrect and plausible because closing C 1100-F034 would be an effective means of increasing the differential pressure necessary to accomplish control rod insertion while venting the CRD overpiston volume and furthermore, the examinee could incorrectly interpret the step at the end of the section that directs opening the C 1100-F034 for any vented control rods not fully inserted as meaning that C11-F034 should be closed when performing this evolution.

Reference Information:29.ESP.03 Section 3.0 (pg. 5)

Plant Procedures 29.ESP.03 NUREG 1123 KA Catalog Rev. 2295037 EK2.Knowledge of the interrelations between SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN and the following:295037 EK2.05CRD hydraulic system 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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0.00 Level of Knowledge:

High Source: BANK: EQOP3150001004 The plant is operating at full power.

Subsequently the following conditions occur:*Main Steamline Radiation Monitors A, B, C, and D all indicate 3,700 mr/hr*Off Gas Radiation Monitor indicates 800 mr/hr increasing* RBHVAC Radiation Monitor indicates 11,000 cpm What is the expected plant response?A.Reactor Scram and a Group 1 Isolation.B.RBHVAC System Isolation and SGTS Initiation.

C.MSIV Isolation, Reactor Scram and Off Gas Isolation.

D.Reactor Scram, RBHVAC System Isolation, and SGTS Initiation.Answer:A ILO 2015 Written Page: 39 of 218 25 June 2015 Answer Explanation:

Reactor Scram and a Group 1 Isolation based on ARP 3D82 Distracter Explanation:

Distacters are incorrect and plausible based on know setpoints list in 3D32, 3D82, & 3D12 Reference Information:

3D82 (pg 1&2) Auto actions and setpoints Plant Procedures 03D082 03D012 03D032 NUREG 1123 KA Catalog Rev. 2295038 EK3Knowledge of the reasons for the followi ng responses as they apply to HIGH OFF-SITE RELEASE RATE:295038 EK3.02System isolations 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

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2.00 Level of Knowledge:

Low Source: NEW EDG 11 is running after a loss of power event. A fire subsequently occurs in the EDG 11 Engine Room, and the CO 2 System has automatically actuated.EDG 11 is (1) and the ventilation system (2)

.A.(1) running (2) fans and dampers are aligned to vent the EDG 11 Engine roomB.(1) tripped (2) fans will shut down, and dampers will isolate the Engine roomC.(1) tripped (2) fans and dampers are aligned to vent the EDG 11 Engine roomD.(1) running (2) fans will shut down, and dampers will isolate the Engine roomAnswer:D ILO 2015 Written Page: 41 of 218 25 June 2015 Answer Explanation:

The CO 2 acutation is not an essential trip for the r unning EDG. The RHR Complex is protected with Wet-pipe Sprinkler, Standpipe Hose Station, and CO 2 Fire Suppression Systems. After CO 2 discharges into an EDG room, the associated ventilation system f ans will shut down, and dampers will isolate the affected room.

Distracter Explanation:

All distracters are plausible if the examinee does not completely understand the interlocks associated with fan operation.A.Is incorrect because the fans and dampers will not align to vent the Engine roomB.Is incorrect because the EDG will not trip.

C.Is incorrect because the EDG wil not trip or fans and dampers align.

Reference Information

23.307 23.501.02 Plant Procedures 23.501.02 23.307 NUREG 1123 KA Catalog Rev. 2600000 AA1.Ability to operate and / or monitor the following as they apply to PLANT FIRE ON SITE:600000 AA1.05Plant and control room ventilation systems.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

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3.00 Level of Knowledge:

High Source: BANK: 315-0028-A020-001 Following a Grid Disturbance, conditions are as follows:

Generator Power 1200 Mwe Reactive Power 360 MVAR (LAG) Generator Hydrogen Pressure 75 psig The System Dispatcher has requested additional reactive load support to maintain grid voltage.

Considering the attached Capablilty Curve, which ONE of the follwing actions is required?

A.RAISE Recirculation Flow to increase the Reactive Load on the Generator.

B.LOWER Recirculation Flow, because Generator Load limits have been exceeded.C.MANUALLY RAISE the Voltage Regulator setting to increase the Reactive Load on the Generator.

D.MANUALLY LOWER the Voltage Regulator setting, because Reactive Load limits have Been exceeded.Answer:C ILO 2015 Written Page: 43 of 218 25 June 2015 Answer Explanation:

RAISING the Voltage Regulator setting in MANUAL will increase the Reactive Load on the Generator.

Is plausible; will raise Generator Power.

A.B.Is plausible; possible misconception regarding 60 psig Hydrogen Curve. Load Limits are exceeded at 60 psig.D.Is plausible; possible misconception regarding 60 psig Hydrogen Curve. Load Limits are exceeded at 60 psig.Plant Procedures 23.118 NUREG 1123 KA Catalog Rev. 2 700000 AA2.01 3.5/3.6 Operating point on the generator capability curve700000 AA2.Ability to determine and/or interpret the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES:

NRC Exam Usage 2015 ILO Exam 2012 ILO Audit Exam ILO 2015 Written Page: 44 of 218 25 June 2015 ID: R21 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW-CONTRACT With Reactor Power starting at 100%, at wh ich of the following points will the first DIRECTED power reduction occur during a Loss of Main Condenser Vacuum event?A.Condenser Vacuum initially starts lowering.B.Initial crew actions fail to stabiliz e Condenser Vacuum between 0.7-2.5 psia.C.Condenser Vacuum reaches 2.8 psia.

D.Condenser Vacuum reaches 3.68 psia.Answer:B ILO 2015 Written Page: 45 of 218 25 June 2015 Answer Explanation:

The first power reduction occurs, due to the crew per forming a Rapid Power Reduction to try to stabilize Condenser Vacuum, when the crew's initial actions were unable to stabilize Condenser Vacuum between 0.7 to 2.5 psia.

Distracter Explanation:

A. Answer is incorrect and plausible because the ex aminee could incorrectly interpret the AOP symptom of 'Main Generator MW decreasing' as meaning R eactor Power would be decreasing. The examinee could also incorrectly choose this response if it is assumed that a Rapid Power Reduction is required immediately upon the start of a lowering Condenser Vacuum event.

C. Answer is incorrect and plausible because the ex aminee could fail to recognize the need to perform a Rapid Power Reduction and instead assume that the first power reduction would occur when the AOP Override condition of 2.8 psia was met requiring t he crew to place the Mode Switch in Shutdown.

D. Answer is incorrect and plausible because the ex aminee because the examinee could fail to recognize the need to perform a Rapid Power Reduction and in stead assume that the first power reduction would occur when the Main Turbine Trip setpoint of 3.68 psia was reached, which would cause a

Reactor Scram due to power being above 30%.

Reference Information:

20.125.01 (pg 3) CONDITION B Plant Procedures 20.125.01 NUREG 1123 KA Catalog Rev. 2295002 AA2.Ability to determine and/or interpret t he following as they apply to LOSS OF MAIN CONDENSER VACUUM :295002 AA2.01Condenser vacuum/absolute pressure.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

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High Source: NEW-CONTRACT Consider each of the transients listed below:*A - Reactor Pressure rises 90 psi above the normal pressure regulator setpoint , then lowers.*B - Reactor Pressure rises to the Reactor Pressure Scram setpoint , then lowers.*C - Reactor Pressure rises 100 psi above the highest SRV safety setpoint , then lowers.*D - Reactor Pressure rises 175 psi above the highest SRV safety setpoint , then lowers.*E - Reactor Pressure rises 225 psi above the highest SRV safety setpoint , then lowers.

Based on the HIGHEST Reactor Pressure encountered dur ing each event, which of these transients would have resulted in the plant NOT being in compliance with (1) LCO 3.4.1.1, Reactor Steam Dome Pressure, and/or (2) SL 2.1.

2, Reactor Coolant System Pressure Safety Limit?A.(1) ALL of the listed transients (2) Transients D and E onlyB.(1) Transients B through E only (2) Transients D and E onlyC.(1) ALL of the listed transients (2) Transients C through E onlyD.(1) Transients B through E only (2) Transients C through E onlyAnswer:B ILO 2015 Written Page: 47 of 218 25 June 2015 Answer Explanation:

By calculation, the maximum pressures encounter ed for each of the transients is as follows:*A - 90 psi above the normal pressure regulator setpoint (944 - 949 psig) is 1034 - 1039 psig.*B - The Reactor Pressure Scram setpoint is 1093 psig.

C - 100 psi above the highest SRV safety setpoint (1155 psig) is 1255 psig.*D - 175 psi above the highest SRV safety setpoint is 1330 psig.

E - 225 psi above the highest SRV safety setpoint is 1380 psig.

(1) LCO 3.4.1.1 pressure is 1045 psig, and (2) SL 2.1.2 pressure is 1325 psig. Only transient A results in pressure <(1), and all of the other transients are > (1). Transients D and E would result in exceeding (2).

Distracter Explanation:

A. Answer is plausible and incorrect because the examinee could incorrectly determine that 100# above the pressure regulator setpoint would result in exceeding 1045 psig.C. Answer is plausible and incorrect because the examinee could incorrectly determine that 100# above the pressure regulator setpoint would result in exceeding 1045 psig and/or the examinee could incorrectly determine that, since transient C resulted in exceeding 1250 psig, which is the design pressure of the RPV suction piping, that the Reac tor Pressure Safety Limit was exceeded which is a common misconception. However, the SL is based on not exceeding 110% of this pressure.

D. Answer is plausible and incorrect because the examinee could incorrectly determine that, since transient C resulted in exceeding 1250 psig, which is the design pressure of the RPV suction piping, that the Reactor Pressure Safety Limit was ex ceeded which is a common misconception. However, the SL is based on not exceeding 110% of this pressure.

Reference Information:

S.L. 2.1

T.S. 3.4.11 NUREG 1123 KA Catalog Rev. 2G2.2.22Knowledge of limiting conditions for operations and safety limits295007High Reactor Pressure Technical Specifications

2.1 SAFETY

LIMITS (SLs)3.4.11 Reactor Steam Dome Pressure 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 48 of 218 25 June 2015 ID: R23 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW-CONTRACT The plant has experienced a steam leak in t he Drywell during operation at rated power. The following conditions currently exist:

RPV water level 197 inchesRPV pressure 800 psig and lowering Reactor power lowering on the SRMs Torus Water Temperature 98°F Drywell Temperature 240°F Drywell Pressure

8.7 psigTorus

Water Level

0.8 inches

What is the basis for placing Torus Spra ys in service for these plant conditions?A.To increase the scrubbing action of the suppression pool while actions to vent primary containment are taking place.B.To cool the enclosed torus airspace by absorbing heat energy through the process of evaporative and convective cooling.C.To lower Drywell Pressure back to within the limits of the Drywell Spray Initiation Limit curve so that actions can be taken to spray the Drywell.D.To effect the desired containment pr essure reduction necessary to draw non-condensable gasses into the Torus airspace as a means of preventing chugging.Answer:B ILO 2015 Written Page: 49 of 218 25 June 2015 Answer Explanation:

The basis for the action to spray the torus is to e ffect the desired pressure drop through cooling of the torus airspace by absorbing heat energy from t he enclosed atmosphere through the process of evaporative and convective cooling. Torus sprays are initiated for the purpose of reducing Primary Containment Pressure and Temperature thus increasing dr ywell cooling. This is the basis for the actions of EOP Steps PCP-4 and PCP-5.A. Is plausible and incorrect because the candidate c ould correctly recall that the Primary Containment is preferentially vented from the torus to take advantage of suppression pool scrubbing for minimizing the amount of radioactivity released. The candidat e could incorrectly conclude that the action to spray the torus is to increase this scrubbing action for containment venting which occurs later in the same EOP parameter control leg at step PCP-13.C.Is plausible and incorrect because the examinee coul d correctly recall that Drywell Pressure must be within the limits of the DWSIL in order to spray the Drywell as directed further down the same EOP parameter control leg at step PCP-8. The exami nee could incorrectly determi ne that spraying of the Torus at step PCP-5 is performed in order to ensure t hat Drywell Pressure is within the limits of the DWSIL when step PCP-8 is reached.D.Is plausible and incorrect because the candidate could incorrectly determine that maintaining non-condensable gasses in the Torus airspace is what prevents the phenomenon of chugging in the Mark I containment. The candidate could incorrectly c onclude that the Before statement of step PCP-4 (Before Torus pressure reaches 9 psig) is meant to lower Torus pressure and therefore draw non-condensable gasses back into the Torus to prevent chugging from occurring, which is a common area

of confusion for licensed operator candidates.

Reference Information:

BWROG EPG APP B (pg B-7-35 TO 37) PCP-4 and PCP-5 Plant Procedures 29.100.01 SH 2 BWROG EPG App B NUREG 1123 KA Catalog Rev. 2295010 AK3Knowledge of the reasons for the follo wing responses as they apply to HIGH DRYWELL PRESSURE:295010 AK3.02Increased drywell cooling.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

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High Source: BANK : ID:41788 Operations is performing a reac tor startup from cold shutdown.

Reactor is critical with a 120 second period.

The Operator withdrawls the next control rod fr om position 08 to 10 which results in a sustained 20 second period.

Which one of the following describes the next Operator action required to be taken?A.Monitor overlap data between SRMs and IRMs and range IRMs as necessary.B.Position SRM detectors as necessary to maintain count rate between 10 2 and 10 5 cps.C.Inform the Reactor Engineer of the power rise, and insert the Control Rod as far as necessary to turn power.D.Insert the Control Rod back to position 08 to obtain a reactor period of > 50 seconds and notify SM and Station Nuclear Engineer.Answer:D ILO 2015 Written Page: 51 of 218 25 June 2015 Answer Explanation:

The rod withdrawal has caused a sustained 20 sec ond period that would result in 3D51 SRM PERIOD SHORT alarm. The alarm response procedure requires the operator to insert control rods to turn power ascension, and notify the SM and SNE.

Distracter Information:

A. This answer is plausible and incorrect if the exam inee were to incorrectly assume that the SRM/IRM overlap verification is required prior to removing SRMs.

B. This answer is plausible and incorrect if the ex aminee incorrectly assumes the period short alarm is caused by the position of the SRMs detectors. This action is warrant ed by the period short alarm ARP but priorty is given to rod insertion.

C. This answer is plausible and incorrect because the first operator action would be to insert the control rod as opposed to inform the Reactor Engineer. The operator would also not be able to insert the control rod as far as necessary because of the RW M preventing the rod from inserting past position 8.

Reference Information:

ARP 3D51(pg 1) SRM PERIOD SHORT Pilgrim Reactivity Event Plant Procedures 03D051 NUREG 1123 KA Catalog Rev. 2295014 AA1.Ability to operate and/or monitor the following as they apply to INADVERTENT REACTIVITY ADDITION:295014 AA1.03RMCS: Plant-Specific.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: NEW-CONTRACT For which of the following conditions would using the Emergency In mode of control rod insertion be authorized without additional Shift Manager concurrence?A.To insert a Control Rod in an attempt to recouple the rod after becoming uncoupled.B.Following a scram where multiple rods failed to fully insert and Reactor Power was 10% on the APRMs and steady.C.Following a scram where several rods failed to fully insert and Reactor Power was on Range 5 of the IRMs and lowering.D.When restoring operation outside of the "E xit" Region of the Power to Flow map following a trip of a Reactor Recirculation Pump with the OPRM Inoperable.Answer:B ILO 2015 Written Page: 53 of 218 25 June 2015 Answer Explanation:

The indications provided are indicative of a Failure to Scram condition where the reactor was still in the heating range, with power above the APRM downscale setpoint of 3%, which requires performance of actions per the Q leg of 29.100.01, Sheet 1A, R PV Control - ATWS. The conditions given would authorize use of 29.ESP.03, Alternate Control R od Insertion Methods, which would then authorize inserting the control rods using t he Emergency In mode of rod insertion.

Distracter Explanation:A. Is plausible and incorrect because the candidate coul d recognize that attempting to recouple a control rod to its mechanism requires inserting the cont rol rod and incorrectly determine that the method used to perform this rod insertion r equires the use of the Emergency In Mode.C. Is plausible and incorrect because the examinee could incorrectly determine that the Emergency In Mode of Rod Insertion is explicitly allowed for this condition without recognizi ng that the rod insertion would NOT be in accordance with the Q leg of the EOPs and instead would be directed by the Scram AOP (due to power being <3%). The candidate could fail to recognize that rod insertion per the Scram AOP is conducted using the normal method of 23.623 and only per ESP 29.ESP.03 with concurrence of the Shift Manager

.D.Is plausible and incorrect because the candidate coul d recognize that operating in the Exit region of the power to flow map, with the OPRM Inoperable, requires control rod insertion to restore operation outside of the Exit region. The candidate could in correctly conclude that t he method of control rod insertion would require use of Emergency In in accordance with the Emergency Power Reduction section of 23.623 without realizi ng that the correct section to use would be for a Rapid Power Reduction that only authorizes use of t he Normal In mode of rod insertion.

Reference Information:

29.100.01 SH 1

29.100.01 SH 1A

29.ESP.03 Plant Procedures 23.623 29.100.01 SH 1A

29.ESP.03 20.106.02 NUREG 1123 KA Catalog Rev. 2295015 AK2.Knowledge of the interrelations bet ween INCOMPLETE SCRAM and the following:295015 AK2.02RMCS: Plant-Specific 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: NEW A condition has resulted in the Spent Fuel Storage Pool Water Level being 1 foot low out of band.

There are no fuel movements or activties ongoing in the Fuel Pool. An LCO entry is required, and there __(1)__ an immediate concern for personnel on the 5 th floor of the Reactor Building because the level requirement is based on __(2)__.A.(1) IS (2) shielding gamma that is produced from decay in an irradiated fuel assembly, while the assembly is being moved above the Reactor CoreB.(1) IS NOT (2) absorbing iodine gases before t hey are released to the secondary containment due to a fuel handli ng accident above the Reactor CoreC.(1) IS (2) shielding neutrons that are produc ed from decay in an irradiated fuel assembly, while the assembly is bei ng moved above the fuel racks in Spent Fuel Storage PoolD.(1) IS NOT (2) absorbing krypton gases before t hey are released to the secondary containment due to a fuel handling a ccident above the Spent Fuel Storage PoolAnswer:B ILO 2015 Written Page: 56 of 218 25 June 2015 Answer Explanation:

LCO 3.7.7 States that The spent fuel storage pool water level shall be

> 22 ft over the top of irradiated fuel assemblies seated in t he spent fuel storage pool racks.

this require per CONDITION ASpent fuel storage pool water level not within limit, immediately suspend movement of irradiated fuel assemblies in the spent fuel storage pool.The bases stat that this is based on absorbing iodine gases before they are released to the secondary containment due to a fuel handling accident above the Reactor Core.

Distracter Explanation:A.This answer is plausible and incorrect. The exam inee would choose this answer if they thought the Spent fuel pool water was meant to shield gamma.C.This answer is plausible and incorrect. The exam inee would choose this answer if they thought the Spent fuel pool water was meant to shield neutrons.D.This answer is plausible and incorrect. The exam inee would choose this answer if they thought the accidents analyzed for were done over the SFP rat her than the core and the gas of concern was krypton.Reference Information:

T.S. 3.7.7 and BASES NUREG 1123 KA Catalog Rev. 2295033 EK1.Knowledge of the operational implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION LEVELS :295033 EK1.02Personnel protection Technical Specifications3.7.7 Spent Fuel Storage Pool Water Level 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (9) Shielding, isolation, and containment design features, including access limitations.

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High Source: NEW An RHR Pump is operating in Shutdown Cooling and the Primary Containment is deinerted when an explosion severely damages equipment in the Primary Containment. Several hours later, the plant has been stabilized.*CHRRMS has been slowing increasing and is currently reading 1041 R/hr.*Primary Containment H 2 concentration has been slowly incr easing and is currently reading 2%.*RBHVAC is shutdown, and Di v 1 SGTS is in service.*A radioactive release is in progress, but ther e is currently no detectable change in dose at the site boundary.

Which of the following procedures would be used to preserve Primary Containment Integrity?A.29.100.01 Sheet 2, Primary Containment ControlB.20.000.02, Abnormal Release Of Radioactive Material C.23.406, Primary Containment Ni trogen Inerting And Purge SystemD.29.ESP.06, Primary Containment V enting And Purge For Hydrogen And Oxygen ControlAnswer:A ILO 2015 Written Page: 58 of 218 25 June 2015 Answer Explanation:

Entry Condition for 29.100.01 Sheet 2 is H 2 >1%.Distracter Explanation:B. Is plausible and incorrect because the examinee thinks that an AOP dealing with the release is the only way to deal with the degradation of primary containment.C. Is plausible and incorrect because the examinee thinks the SOP can be use to purge H 2 and therefore help preserve Primary Containment.

D. Is plausible and incorrect because th is is the ESP that the EOPs would direct if venting were allowed, but for this condition, venting for H 2 control is not allowed.

Reference Information:

29.100.01 Sheet 2 - Entry conditions Plant Procedures 29.100.01 SH 2 NUREG 1123 KA Catalog Rev. 2500000 EK1.Knowledge of the operational implications of the following concepts as they apply to HIGH CONTAINMENT HY DROGEN CONCENTRATIONS:500000 EK1.01Containment integrity 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.

NRC Exam Usage 2015 ILO Exam Early Review Questions for NRC 2015 ILO ILO 2015 Written Page: 59 of 218 25 June 2015 ID: R28 Points: 1.00Difficulty:

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High Source: NEW-CONTRACTThe plant was operating at 100% power when a small primary coolant leak raised drywell pressure to 2.2 psig. All systems operated as designed.

The following conditions exist when the plant is initially stabilized:*RPV level is 195 inches being maintained using RCIC.*RPV pressure is 900 psig.

Division 1 RHR is in Torus Cooling Mode usi ng 'A' RHR pump; 'A' RHR CMC switch is in RUN.*A' RHR Pump MANUAL OVERRIDE white light is illuminated.

All other RHR Pumps are running with CMC switches in AUTO.

A loss of offsite power then occurs.

Select the response that describes the operati on of the RHR pumps as the EDGs automatically load with NO operator action.A.All RHR pumps start immediately when t heir respective EDG's output breaker closes.B.All RHR pumps start 5 seconds after t heir respective EDG's output breaker closes.C.RHR pumps 'B', 'C', and 'D' start imm ediately when their respective EDG's output breaker closes. 'A' RHR pump is stopped.D.RHR pumps 'B', 'C', and 'D' start 5 seconds after their respective EDG's output breaker closes. 'A' RHR pump is stopped.Answer:A ILO 2015 Written Page: 60 of 218 25 June 2015 Answer Explanation:

The design of the EDG load sequencer starts RHR pumps immediately with no time delay. Also, the A RHR pump CMC in RUN will not prevent the pump from restarting after having been taken out of AUTO during and ECCS initiation. If the CMC was returned to AUTO after being taken out of AUTO after the ECCS initiation signal was present the pump would remain OFF.

Distracter Explanation:B.Is incorrect and plausible. The examinee could in correctly identify that the RHR pumps have a timed delay for starting as many components initiated by the EDG load sequencer have associated time delays.C.Is incorrect and plausible. The examinee could confuse the MANUAL OVE RRIDE circuit as being applicable in the AUTO and RUN positions of the CMC switch since the light is illuminated with the CMC in RUN with an ECCS initiation signal present. If the CMC was returned to AUTO after being taken out of AUTO after the ECCS initiation signal was present the pump would remain OFF.D.Is incorrect and plausible. The examinee could in correctly identify that the RHR pumps have a timed delay for starting as many components initiated by the EDG load sequencer have associated time delays. The examinee could confuse the MANUAL OVERRIDE circ uit as being applicable in the AUTO and RUN positions of the CMC switch since t he light is illuminated with the CMC in RUN with an ECCS initiation signal present. If the CMC was re turned to AUTO after being taken out of AUTO after the ECCS initiation signal was present the pump would remain OFF.

Reference Information:

I-2714-24 EDG AUTO LOAD SEQ (H-2) RHR is before step 1 SO start at time 0. (drawing in development folder)I-2201-01 RHR PUMP A ELECTRICAL SCHEMATIC (E-3 thru H-3) 52XX and K9A and XK-33 (drawing in development folder)

Plant Procedures 23.205 NUREG 1123 KA Catalog Rev. 2 203000 RHR/LPCI: Injection Mode203000 K4.Knowledge of RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following:203000 K4.07Emergency generator load sequencing NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 61 of 218 25 June 2015 ID: R29 Points: 1.00Difficulty:

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High Source: BANK 315-0141-A021-029 The plant is shutdown and the following conditions exist:*BOTH Recirc Pumps are secured. *Shutdown Cooling is being provided by RHR Loop A.

Reactor Coolant System Temperature is 170°F, with a cooldown in progress. *RPV Water Level LOWERED to 190 inches on the Narrow Range Level indicators.

RPV Water Level is a concern because it is __________________________.A.too low to prevent RHR Pump cavitation.B.too low to prevent RPV thermal stratification.

C.low enough to generate a Low RPV Level scram signal.

D.low enough to generate a RHR Shutdown Cooling Isolation signal.Answer:B ILO 2015 Written Page: 62 of 218 25 June 2015 Answer Explanation:

Per 23.205 3.2.2 During Non-ATWS, RHR SDC Mode operat ion, reactor water level must be maintained above 220 inches to prevent temperature stratificati on. During ATWS, RPV level is controlled per EOPs.

Distracter Information:A. Is plausible and incorrect because RWL in this case provides head to the RHR pump.C. Is plausible and incorrect because there is a low RPV level scram

D. Is plausible and incorrect because ther e is a low RPV level Isolation for SDC.

Reference Information:

23.205 (pg 10) P&L 3.2.2 Plant Procedures 23.205 NUREG 1123 KA Catalog Rev. 2205000 K5.Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) :205000 K5.03Heat removal mechanisms 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

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Low Source: NEW-CONTRACT With the plant operating at full power, a failure of the CST level instrumentation has caused HPCI suction to swap to the Torus.

HPCI should be considered INOPERABLE if ___________________________________.A.this condition is maintained for more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B.this condition is maintained for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C.Keep Fill is also lost in this condition for more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D.Keep Fill is also lost in this condition for more than 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months sAnswer:C ILO 2015 Written Page: 64 of 218 25 June 2015 Answer Explanation:

The SOP has a P&L that requires that ff aligned to t he Torus in standby for more than twelve consecutive hours without HPCI Keep Fill Syst em in operation, HPCI should be considered INOPERABLE, due to potential drain down of system piping.

Distracter Explanation:A. Is plausible since HPCI is not aligned to its nor mal suction source, the CST. The examinee could fail to recognize that the transient and accident analyses , which take credit for HPCI, assume that the HPCI suction source is the suppression pool.B.Is incorrect but plausible because the HPCI sy stem piping could become drained if it remained aligned to the Torus for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , requiring it to be declared INOPERABLE. D.Is plausible because the combination of HPCI suction being aligned to the Torus with a loss of Keep Fill for more than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months would cause HPCI to be INOPERABLE, but HP CI is INOPERABLE at more than 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Reference Information:

23.202 (pg 7) P&L 3.19 Plant Procedures 23.202 NUREG 1123 KA Catalog Rev. 2206000 K6.Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE COOLANT INJECTION SYSTEM :206000 K6.09Condensate storage and transfer system: BWR-2,3,4 Technical Specifications3.3.5.1 Emergency Core Cooling System (ECCS) Instrumentation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: BANK: 315-0140-B003-001 A LOCA signal from Div 1 CS Logic is received while DIV 1 Core Spray is operating in TEST Mode. How will the Core Spray System respond?A.Div 1 Core Spray will trip. Div 2 Core Spray will line up to inject.

B.Div 1 and 2 Core Spray will line up to inject. The test valve will close.

C.Div 1 Core Spray will remain running in the TEST mode. Div 2 Core Spray will line up to inject.D.Div 1 Core Spray will remain running in the TEST mode. Div 2 Core Spray will remain in Standby.Answer:B ILO 2015 Written Page: 66 of 218 25 June 2015 Answer Explanation:

A LOCA signal from either CS Logic will auto start both divisions of Core Spray.

Per 23.203, Section 5.5, Auto Initiation Div 1, "If Di v 1 Core Spray is in the Test Mode when an automatic initiation signal is received, E2150-F015A, Div 1 CS Test Line Iso Vlv, will close as the system aligns for injection to the Reactor Vessel." Distracter Explanation:A.Is incorrect and plausible if examinee incorrectly assumes that system lineup would cause a trip condition when an injection signal occurs.C.Is incorrect and plausible if exami nee did not know about the auto closure.D.Is incorect and plausible if the examinee felt that a LOCA signal from only Div 1 CS Logic does not provide a system start signal.

Reference Information:

23.203 Section 5.5 (pg 22) NOTE Plant Procedures 23.203 NUREG 1123 KA Catalog Rev. 2209001 A1.Ability to predict and/or monitor changes in parameters associated with operating the LOW PRESSURE CORE SPRAY SYSTEM controls including:209001 A1.08System lineup 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

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High Source: BANK: 315-0114-C010-006 The SLC Initiation Keylock Switch, C4100- M004, has been placed in the PMP A RUN position.

The following indications are noted 30 seconds later: *Reactor Pressure is 1000 psig. *C41-R601, SLC Tank Level Indicator, is steady.

SLC Continuity Lights A and B are ON.

SLC Pump A CMC Switch red light is ON, and green light is OFF.

C41-R600, SLC Pump Discharge Pressure Indicator, is oscillating between 1320 and 1370 psig.These are indications of (1) what condition, and (2) what action should the operator perform?A.(1) SLC Explosive Valves failed to fire.

(2) Start SLC Pump B.B.(1) Normal operation for the SLC System.(2) Monitor SLC Tank level.C.(1) C41-F001, SLC Storage Tank Outlet Valve is closed.

(2) Dispatch an operator to open C41-F001.D.(1) C41-F029A, SLC Pump A Discharge Relief Valve failed open.

(2) Dispatch an operator to gag shut C41-F029A.Answer:A ILO 2015 Written Page: 68 of 218 25 June 2015 Answer Explanation:

If the C41-F004A & B, failed to fire, positive di splacement SLC Pump A will OPEN C41-F029A, SLC Pump A Discharge Relief Valve, which causes pr essure oscillations between 1320 and 1370 psig. These are the lift and reseat pressures for C41-F029A. Star ting SLC Pump B will fire the other primer in both valves.Distracter Explanation:B.Is incorrect and plausible, normal Indication would be discharge pressure slightly higher than Reactor Pressure AND lowering SLC Tank Level.C.Is incorrect and plausible, the Tank Level would re main steady if the Storage Tank Outlet were shut, but discharge pressure would be low.D.Is incorrect and plausible, Relief Valve has opened, but has not failed.

Reference Information:

SOP 23.139 (Pg 11&12) SLC injection Plant Procedures 23.139 NUREG 1123 KA Catalog Rev. 2211000 A2.Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM ; and (b) based on those predictions, us e procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:211000 A2.06Valve openings 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

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High Source: MODIFIED: 315-0127-C010-001 The plant is shut down in a refueling out age with undervessel work in progress on SRM B detector. The restoration sequence of the "Mode Switch in Refuel and One Rod-Out Interlock Verification" surveillance is in progress.

The Reactor Mode Switch is placed in SHUTDOWN. The Scram Reset Switch is then turned to the GP 1/4 AND GP 2/3 positions, and released. About one minute later, an automatic scram si gnal is received. All RPV and Containment parameters remain constant throughout the event.

Assuming no other operator actions were performed, which of the following explains the cause of the SECOND scram?A.Alarm 3D51, SRM PERIOD SHORT, was received due to moving the SRM detector.B.Alarm 3D56, TESTABILITY LOGIC A/

B RPS/PWR FAILURE, was received due to a blown fuse in RPS Cabinet H21-P085.C.Alarm 3D86, MN STM LINE ISO VALVE CLOSURE CHANNEL TRIP, was received due to an upscale failure of a Main Steam Line Flow instrument.D.Alarm 3D94, DISCH WATER VOL HI LEVEL CHANNEL TRIP, was received due to the SDV High Level Channel Trip not being bypassed before the first scram was reset.Answer:D ILO 2015 Written Page: 70 of 218 25 June 2015 Answer Explanation:

SDV High Level will initiate a second automatic reac tor scram under the given conditions. This question is based on OE at Fermi (LER 96-021-00) involving failure to bypass the SDV High Level channel trip prior to reseting a scram.

Distacter Explanation:A.Is plausible and incorrect; this alarm could be ex pected during SRM insertion, but is ONLY an alarm.B.Is plausible and incorrect; RPS power faillure in an RPS cabinet could cause an alarm, but not a scram.C.Is plausible and incorrect; but the MSIV Closure Tr ip is bypassed with the Reactor Mode Switch in SHUTDOWN.Reference Information:

ARP 3D94 SOP 23.610 pg 10&11

LER 96-021 Plant Procedures 23.610 03D094 Operating Experience LER 96-021 Fermi Auto Scram on SDV during Shutdown NUREG 1123 KA Catalog Rev. 2212000 A3.Ability to monitor automatic operations of the REACTOR PROTECTION SYSTEM including:212000 A3.05SCRAM instrument volume level 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: NEW-CONTRACT A reactor startup is in progress in accordanc e with 22.000.02, Plant Startup to 25% Power. Reactor power is 7% as indicated on the Average Power Range Monitors (APRMs). Source range detectors have been withdrawn. Intermediate range detectors are being withdrawn. Intermediate range monitors (IRM) A, B, C, and D indicate 20 on range 9. IRMs E, F, G, and H indicate 30 on range 10. During withdrawal of IRM F, the retract per mit logic malfunctions and the RETRACT PERMIT light extinguishes. Based on these conditions, whic h of the following describes the IRM system and related system response?A.IRM F stops retracting, and 3D113, CONTROL ROD WITHDRAWAL BLOCK, alarms.B.IRM F stops retracting, and 3D60, IRM CH B/F/D/H UPSCALE TRIP/INOP, alarms causing a half scram.C.IRM F continues to retract, and 3D113, CONTROL ROD WITHDRAWAL BLOCK, alarms.D.IRM F continues to retract, and 3D60, IRM CH B/F/D/H UPSCALE TRIP/INOP, alarms causing a half scram.Answer:C ILO 2015 Written Page: 72 of 218 25 June 2015 Answer Explanation:

The IRM RETRACT PERMIT light should be on when the Mode Switch is placed in RUN. In accordance with procedure 22.000.02 , Plant Startup to 25% Power, IRMs are not withdrawn until the Mode Switch is placed in RUN which is directed at 5%-10% power.

The IRM RETRACT PERMIT interlock causes a rod block if the IRM is withdrawn and the interlock is not satisfied (light extinguished). The interlock does not prevent continued withdr awal of the detector.

Distracter Information:A.Is incorrect and plausible. Misapplying or mi sunderstanding the IRM RETRACT PERMIT logic would allow the examinee to select this answer, also it is plausible that a malfunction of the IRM system during a startup would cause a rod blockB.Is incorrect and plausible. Misapplying or mi sunderstanding the IRM RETRACT PERMIT logic would allow the examinee to select this answer, also it is plausible a malfunction of the IRM system would cause an IRM INOP trip. The INOP trip does cause a half scram.D.Is incorrect and plausible. The IRM does continue to retract, several malfunc tions cause an IRM INOP trip which does cause a half scram if alarmed.

Misapplying the IRM INOP for a malfunction in the RETRACT PERMIT logic would allow the examinee to se lect this answer. The INOP trip does cause a half scram.

Reference Information:

I-2115-6 Shows how mode switch effects rod blocks from SRM/IRM

I-2145-56 Show IRM retact permit logic

I-2145-59 Shows all rod blocks from IRMs Plant Procedures 23.603 22.000.02 NUREG 1123 KA Catalog Rev. 2 215003 IRM System 215003 K1.Knowledge of the physical connections and/or cause effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following:215003 K1.02Reactor manual control NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 73 of 218 25 June 2015 ID: R35 Points: 1.00Difficulty:

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High Source: NEW With the Reactor Mode Switch in STARTUP HO T STANDBY, the Intermediate Range Monitors are on RANGE 4 with the following readings:

IRM Channel Reading A 103 B 101 C 102 D 102 E 108 F 95 G 103 H 98 The P603 operator places IRM Channel E Range Selector Switch in RANGE 3.Which of the following conditions are the result of the P603 operator's action?A.IRM Downscale Alarm ONLYB.IRM Upscale Trip ONLYC.IRM Upscale Trip AND Control Rod Withdrawal Block ONLYD.IRM Upscale Trip AND Control Rod Withdrawal Block AND Half-ScramAnswer:D ILO 2015 Written Page: 74 of 218 25 June 2015 Answer Explanation:

IRM E is just below the UPSCAL E TRIP. Ranging DOWN will result in IRM E causing an UPSCALE TRIP, ROD BLOCK, and a Half Scram.

Distracter Explanation:

A. Is plausible and incorrect an incorrect assumpti on that down-ranging causes a downscale condition. B. Is plausible and incorrect an incorrect assumption that down-ranging causes an upscale trip only.

C. Is plausible and incorrect an incorrect assumpti on that down-ranging causes an upscale trip and rod block with no half-scram.

Reference Information:

ARP 3D113, 3D59 and 3D60 Plant Procedures 03D059 03D060 03D113 NUREG 1123 KA Catalog Rev. 2215003 A4.Ability to manually operate and/or monitor in the control room:215003 A4.03IRM range switches 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: MODIFIED BANK: 3150122A013005 During a reactor startup, reactor power is on Range 3 of the Intermediate Range Monitors, and Source Range Monitor (SRM) detectors are being withdrawn from the core.

DIV 2 48/24V DC Distribution loses power.

Assuming no operator action is taken, a Contro l Rod Block is caused by power loss to _____ .A.SRM A and SRM D ONLY.

B.SRM B and SRM C ONLY.

C.SRM A,C and IRMs A,C,E,G,I.

D.SRMs B, D and IRMs B,D,F,H.Answer:D ILO 2015 Written Page: 76 of 218 25 June 2015 Answer Explanation:Power to IRM Channels B, D, F, H and SRM Channels B and D are from 48/24V DC Distribution Cabinet 2IB1-3, Circuit Breaker 1, Distracter Explanation:

Distracters are plausible as SRM drives are power ed from 120/208V Distribution Cabinet 72E-2B-1 vice 48/24V and knowing the ROD BLOCK setpoint.

References Information:

ARP 3D113, ROD BLOCK 300 V DC decreasing (pg. 2), SOP 23.602, ENERGIZING SRM (Pg 9).

Plant Procedures 03D113 23.602 NUREG 1123 KA Catalog Rev. 2215004 K2.Knowledge of electrical power supplies to the following:215004 K2.01SRM channels/detectors 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: NEW-CONTRACT Which of the following statements is an accurate description of the operation of the Source Range Monitoring System regarding simultaneous movement of all four SRM detectors?A.It is NOT possible to do so at any time due to limitations imposed by the control circuitry.B.It is NOT permissible to do so at any ti me as this will result in unreliable indications as the detector travels through flux in the core.C.It is ONLY possible to do so in the inward direction during a Reactor Shutdown due to limitations imposed by the control circuitry.D.It is NOT permissible to do so during a R eactor Startup as this will result in unreliable indications as the detecto r travels through flux in the core.Answer:D ILO 2015 Written Page: 78 of 218 25 June 2015 Answer Explanation:

The examinee should correctly determi ne that simultaneous detector movement is always allowed by the detector control circuitry; however, 23.602 P&L 3.11 prohibits simultaneous detector movement during a Reactor Startup as this will result in an indicat ed change in log count rate and period as the detector travels through flux in the core.

Distracter Explanation:A. Is incorrect and plausible because the candidat e could incorrectly determine that simultaneous detector movement is never allowed and enforc ed by programming within the detector control circuitry. This is also made plausible by the fact that most of the P&Ls in 23.602 cover system limitations, response, etc., impos ed by the SRM control circuits.B. Is incorrect but plausible because the exam inee could incorrectly det ermine that simultaneous movement of SRM detectors is never permissible due to unreliable indications, when in fact the P&L only prohibits simultaneous movement during Reactor Startup.C. Is plausible because the candidate could remember that it is only permissible to simultaneously insert SRM detectors following a Reactor Scram and incorre ctly determine that this limitation must be imposed by the control circuitry. This is also m ade plausible by the fact that most of the P&Ls in 23.602 cover system limitations, response, etc., imposed by the SRM control circuits.

Reference Information:

23.602 (pg 8) P&L 3.11 Plant Procedures 23.602 NUREG 1123 KA Catalog Rev. 2G2.1.32Ability to explain and apply system limits and precautions215004SRM System 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

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High Source: BANK: 315-124-0008-005Which of the following describes a set of conditions that would generate a full reactor scram signal?A.APRMs 1 & 4 failed downscale during normal operations.B.APRM 4 becomes inop during refueling with the shorting links installed.

C.APRMs 2 & 3 failed upscale during a startup with the mode switch in RUN.

D.APRM 1 failed upscale with IRM A failed upscale during a startup with the mode switch in STARTUP/HOT STANDBY.Answer:C ILO 2015 Written Page: 80 of 218 25 June 2015 Answer Explanation:

Per 23.601 INSTRUMENT TRIP SHEETS Any 2 APRMs in a tripped state will cause the 2 of 4 vo ters to initiate a full Rx Scram & Neutron Flux -

Upscale (Setdown) <

15% RTP Distracter Information:

A.Is plausible and incorrect, bec ause 2 APRMs have input to vote rs which would normally be enough votes to cause a scram, however downscale will not cause a vote.B.Is plausible and incorrect, bec ause an APRM has input to vote rs which can cause a scram and shorting links normally reduce coincidence, however only 1 APRM INOP will not vote a scram even with shorting links installed.D.Is plausible and incorrect, because 2 votes causes a scram, however the IRM does not input into the voters.Reference Information:

23.601 (pg 41) Average Power Range Monitor (Scr am setpoints, input to 2 of 4 voter)

Plant Procedures 23.601 NUREG 1123 KA Catalog Rev. 2215005 K1.Knowledge of the physical connections and/or cause effect relationships between AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM and the following:215005 K1.01RPS 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (6) Design, components, and func tion of reactivity control mechanisms and instrumentation.

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High Source: NEW-CONTRACTA plant startup is in progress. Reactor power is 30% and being increased by control rod withdrawal. APRM 2 malfunctions and is indicating 0% power.

Which one of the following describes the effect on Rod Block Monitors as a result of the APRM failure?A.RBM A and B are OPERABLE. Outwar d Control rod motion can continue.B.RBM A and B are INOPERABLE. Outwar d Control rod motion is blocked.C.RBM A is bypassed and INOPERABLE. Bypassing APRM 2 will allow outward control rod motion.D.RBM B is bypassed and INOPERABLE. Bypassing APRM 2 will allow outward control rod motion.Answer:D ILO 2015 Written Page: 82 of 218 25 June 2015 Answer Explanation:

The examinee should correctly recognize that APR M #2 is the primary APRM for RBM B and determine that, under the current condition of 30% power, the RBM is bypassed due to APRM 2 power input of 0%.

The RBM is bypassed until power exceeds the setpoi nt of 27% power from the primary APRM. The secondary ARPM (APRM 4 in this case) will be automat ically selected after the ARPM 2 bypass joystick is operated.This question is made more challenging by the fact that Tech Specs requires the RBM to be OPERABLE when Power is 30%, while the RBM actually starts enfor cing when Power is >27% by the reference APRM.Distracter Explanation:A.Is incorrect and plausible because the examinee may incorrectly recall that technical specification allowable setpoint for RBM bypassing is 30% and determine that the RBM are not yet affected. Additionally RBM A is associated with APRMs #1 and 3, and since it is may seem more logical to associate APRMs 1&2 with RBM A, rather than APR Ms 2&4 being associated with RBM B. Also with the APRM downscale with the mode switch in run, outward rod motion could not occur due to a

APRM downscale control rod withdrawl block.B.Is incorrect but plausible because the examinee could incorrectly determine that the ARPM downscale effects both RBMs. (Recirc flow trans mitter failures could affect both RBMs flow comparator). Outward control rod motion is blocked by the APRM downscale failure.C. Is incorrect but plausible because the examinee ma y incorrectly recall that RBM A is associated with APRM #2 since it is may seem more logical to associate APRMs 1&2 with RBM A, rather than APRMs 2&4 being associated with RBM B.

Reference Information:

23.607 (pg 3) 1.1 System Description ILO 2015 Written Page: 83 of 218 25 June 2015 Plant Procedures 23.605 03D099 03D103 23.607 NUREG 1123 KA Catalog Rev. 2215005 K3.Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following:215005 K3.07Rod block monitor: Plant-Specific Technical Specifications3.3.2.1 Control Rod Block Instrumentation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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Low Source: NEW-CONTRACT Which of the following RCIC System Trips/Isolat ion parameters is designed with the purpose of providing protection for the RCIC Turbine?A.Steam Line FlowB.Pump Suction Pressure C.Exhaust Diaphragm Pressure D.Equipment Room TemperatureAnswer:C ILO 2015 Written Page: 85 of 218 25 June 2015 Answer Explanation:

The examinee should correctly determine that High tu rbine exhaust diaphragm pressure indicates that the pressure may be too high to continue operation of the RCIC system's turbine. The candidate should recognize that, when this isolation setpoint is reached, one of two exhaust diaphragms has ruptured and pressure is reaching turbine casing pressure lim its. Furthermore, the c andidate should understand that these isolations are for equipment protection and are not assumed in any transient or accident analysis in the UFSAR.

Distracter Explanation:A.Is plausible because the Steam Line Flow High f unction is provided to detect a break of the RCIC steam lines and initiate closure of the steam line isolation valves of the RCIC system. The candidate could incorrectly conclude that the Steam Line Fl ow High function protects the RCIC Turbine from either high steam flow induced damage, such as over speed. This is possible if the candidate failed to recognize that, if the steam is allowed to c ontinue flowing out of the break, the reactor will depressurize and the core can uncover, therefore, the isolations are initiated on high flow to prevent or minimize core damage.B.Is plausible because the Pump Suction Pressure is provided to protect the pump against possible cavitation and lack of cooling. The candidate could in correctly conclude that the Pump Suction Trip protects the turbine by tripping the pump. D.Is plausible because area temperat ures are provided to detect a l eak from the associated system steam piping and an isolation of t he RCIC steam supply valves occurs on a sensed high temperature condition. The candidate could incorrectly concl ude that the High Equipment Room Temperature function protects equipment in the RCIC room. The c andidate could also incorrectly conclude that this choice is possible if the candidate remembered t hat this function is diverse to the high flow instrumentation, as described in the Technical Spec ification Bases, but failed to recognize that the isolation occurs when a very small leak has occu rred because, if the small leak were allowed to continue without isolation, o ffsite dose limits may be reached.

Reference Information:

T.S.B 3.3.6.1 (pg 3.3.6.

1-16&17) RCIC Turbine Exhaust Diaphragm Pressure-High ILO 2015 Written Page: 86 of 218 25 June 2015 NUREG 1123 KA Catalog Rev. 2217000 K4.Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design feature(s) and/or interlocks which provide for the following:217000 K4.04Prevents turbine damage: Plant-Specific Technical Specifications3.3.6.1 Primary Containment Isolation Instrumentation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: BANK: 315-0162-0005-001What is the effect on the ADS System if UPS power is lost?A.ADS Level 1 Logic deenergized.B.ADS Level 3 Logic deenergized.

C.ADS Logic A and B deenergized.

D.ADS Control Room Timer deenergized.Answer:D ILO 2015 Written Page: 88 of 218 25 June 2015 Answer Explanation:Per 3D22 Loss of UPS A for ADS Timers Distracter Explanation:A.Is incorrect and plausable Level 1 is an input to ADSB.Is incorrect and plausable Level 3 is an input to ADS C.Is incorrect and plausable A/B logic triggers ADS.

Reference Information:

ARP 3D22 (pg 1)

ILO 2015 Written Page: 89 of 218 25 June 2015 ILO 2015 Written Page: 90 of 218 25 June 2015 Plant Procedures 03D022 NUREG 1123 KA Catalog Rev. 2218000 K2.Knowledge of electrical power supplies to the following:218000 K2.01ADS logic 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: NEW-CONTRACT A transient has occurred resulting in the following conditions: *ALL MSIVs are closed. *No operator actions have been taken.

RCIC is operating and injecting into the vessel.

RPV Water Level is 35 inches, lowering one inch per minute.

Reactor pressure is 900 psig and slowly lowering 5 psig per minute.

Drywell pressure is 1.0 psig and trending up at 0.05 psig per minute.

1D57, ADS/SRV/EECW TCV POWER SUPPLY FAILURE, is in alarm. *2PA2-5 Circuit 1 is de-energized.

What effect will the above conditions have on ADS if NO operator actions are taken within the next 15 minutes?A.The ADS valves will NOT have been opened when required by the EOPs thereby challenging fuel design limit s due to loss of adequate core cooling.B.The ADS valves will have automatica lly opened before being required to do so by the EOPs thereby complicating effo rts to restore and maintain RPV water level.C.Only three of the ADS valves will have opened automatically and therefore the RPV is NOT ensured to remain depressurized under all conditions as required by the EOPs.D.Power to the ADS valves will not hav e been restored, thereby rendering them incapable of opening automatically or re motely by the operator and removing the blowdown function of the ADS valves.Answer:B ILO 2015 Written Page: 92 of 218 25 June 2015 Answer Explanation:

The examinee should correctly determine that in 3 minutes, Level 1 (31.8") will be reached, and 7 minutes after that the 105 second timer will count down after which the ADS valves will open. The examinee should conclude that, within the 15 minutes with no operator action tak en, the ADS valves will be open. The examinee should also determine that ADS Logic String A has lost power; however, ADS Logic String B still has power from an alternate source (2PA2-6, Circuit 1) and that the power source supplying this logic string also powers the pilot valve's solenoids. The examinee should determine that the EOPs require ADS to be inhibited at Level 1, and failure to do will result in an uncontrolled ADS initiation at 20" RPV level, which is before the EOP required point of WH EN RPV level is = 0".

Distracter Explanation:A. Is plausible because the candidate could calculat e that the ADS valves should open automatically in 10 minutes and are prevented from doing so due to t he logic power failure and therefore incorrectly conclude that, with no actions taken at that point and the ADS valves not being opened, that the crew is in violation of the EOPs and their bases. C.Is plausible because the ADS valves will open in 10 minutes and will be open at the 15 minute point and the candidate could incorrectly determine that the two ADS logic power supplies go to different ADS valves/valve logic strings thereby resulti ng in less than the EPG defined Minimum Number of SRVs Required for Emergency Depressurization (MNSRE D) to be open, the definition of which is the second half of this distractor.D. Is incorrect but plausible because 2PA2-5 Circuit 1 is the normal power suppl y to the ADS pilot valve solenoids and the examinee could incorrectly determi ne that, since this power source was lost and no subsequent action taken, that the ADS valves would be incapable of performing their required function either automatically or manually as required by the EOPs.

Reference Information:

ARP 1D31 indicates that ADS will time down for 7 minutes before initiating the timer with only a L1 (31.8"). The high drywell pressure signal is not presen t, and will not be due to the rate of pressure rise as given in the stem of the question.

ARP 1D44 indicates that after initia tion ADS will time down for 105 seconds.

ARP 1D36 indicates that ADS valves will open after the timer is complete.ARP 1D57 validates that some ADS components are de-energized due to 2PA2-5 Ckt 1 off as indicated in the stem The logic prints for ADS power supplies (I-2095-01) and B ADS logic power (I-2095-07) indicate how the logic remains automatically powered on a loss of the power from 2PA2-5 Ckt 1. (located in drawings.pdf)

The logic prints for ADS valves P, J, R (I-2095-02) indicate how the valves remains automatically powered on a loss of the power from 2PA2-5 Ckt

1. Valves H and E are the same. (located in drawings.pdf)

The BWROG bases for step RC/L-2 describes that ADS should be prevented as follows:

Reference Discussion :

If it has been determined that RPV water level can be restored and maintained above the top of the active fuel with available injection sources, emer gency RPV depressurization is unnecessary. Automatic initiation of ADS is therefore pr evented. Subsequent steps pr ovide explicit and detailed instructions for controlling RPV water level and pressure and spec ify when emergency depressurization is appropriate.

Permitting automatic ADS initiation ma y be undesirable for the following reasons:

ILO 2015 Written Page: 93 of 218 25 June 2015

ADS actuation can impose a severe thermal transi ent on the RPV and may complicate efforts to control RPV water level.

If only steam-driven systems are available for inject ion, ADS actuation may directly lead to loss of adequate core cooling and subsequent core damage.

The conditions assumed in the design of the ADS ac tuation logic (e.g., no operator action for 10 minutes after event initiation) may not exist when the ac tions specified in this step are being performed.

The operating crew can draw on much more inform ation than is available to the ADS logic (e.g., equipment out of service for main tenance, operating experience with certain systems, probability of restoration of off-site power, etc.) and can better judge, based on instructions contained in the EPGs/SAGs, when and how to depressurize the RPV.

ADS initiation is prevented in this step, however, only if RPV water level can be restored and maintained above the top of the active fuel and the ADS timer initiates. If RPV water level cannot be restored and maintained above the top of the active fuel, the instru ction is not applicable, sinc e level control transfers to Contingency #1 in accordance with the branch at t he end of Step RC/L-2. If the timer does not initiate, the ADS logic is not defeated so that the system will still provide an automatic backup for high pressure injection in small break loss of coolant accidents.

Plant Procedures 01D31 01D36 01D44 01D57 BWROG EPG App B NUREG 1123 KA Catalog Rev. 2218000 K5.Knowledge of the operational implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION SYSTEM:218000 K5.01ADS logic operation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

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High Source: BANK: EQOP3150165A016002 Following a transient initiated by a loss of the 345 kV and 120kV switchyards, the following Emergency Diesel Generat or conditions exist: *EDG 11 Jacket Coolant Temperature High is alarming. *EDG 12 Crankcase Pressure High is alarming.

EDG 13 Oil Temperature High is alarming.

EDG 14 Fuel Oil Pressure Low is alarming.

Which one of the Core Spray Pumps is affected?A.Core Spray Pump A B.Core Spray Pump B C.Core Spray Pump C D.Core Spray Pump DAnswer:C ILO 2015 Written Page: 95 of 218 25 June 2015 Answer Explanation:

With a Loss of Offsite Power, Undervoltage conditi ons have caused EDGs to start with bypassed non essential trips. Crankcase Pressure High is an Ess ential Trip and EDG 12 is TR IPPED, this is the power source to Core Spray Pump C.

Distracter Explanation:A.Is plausible; would be true if Jacket Coolant Tem perature High were an Essential Trip and Crankcase Pressure High was a Non Essential Trip.B.Is plausible; would be true if Oil Temperature Hi gh were an Essential Trip and Crankcase Pressure High was a Non Essential Trip.D. Is plausible; would be true if Fuel Oil Pressure Low were an Essential Trip and Crankcase Pressure High was a Non Essential Trip.

Reference Information:

SOP 23.307 Section 1.1 page 5 behavoir of EDG with start signal & Encl B for trips Plant Procedures 23.307 NUREG 1123 KA Catalog Rev. 2264000 K3.Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following:264000 K3.01Emergency core cooling systems 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: BANK: 315-0148-0005-007 The plant is operating normally at 95% power.

A significant leak develops on the H21-P004 rack, and the excess flow check valve on the VARIABLE leg of the WIDE Range instrument s closes in response to the leak.

What would DIRECTLY be the effect on t he plant from this change in sensed level?A.RWCU isolates.

B.TIPs retract and isolate.

C.Reactor low water Level 3 scram.

D.Level 8 trip of both RFPs and Main Turbine.Answer:A ILO 2015 Written Page: 97 of 218 25 June 2015 Answer Explanation:

Wide range instruments on a rack means that the l eak affects 2 instruments, and because the check valve is closed, the variable leg of the two instrument will be low, so that the instruments will indicate failed low. Because the rack number is even P 004, and the logic is NSSS A= Channels A, B and B=Channels C,D. Therefore NSSS group isolations for the wide range instruments would occur on NSSS logic A. In this case Group 10 for RWCU is one isolations that would occur Distracter Explanation:B.Is plausible and if student thinks group 15 is off of wide range or connects this leak to a narrow level instruments.C.Is plausible if students thinks there is a RPS Tr ip associated with the wide range instruments or connects the leak a narrow range instrument,D.is plausible if the examinee believes this leak affe cts narrow range or associates the Level 8 trip with wide range.

Reference Information:

SOP 23.601 (Pg 11-12) Instrument trip sheets FOS M-5701-2 Instrument location.

Plant Procedures 23.601 NUREG 1123 KA Catalog Rev. 2223002 K6.Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT ISOLATIO N SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF:223002 K6.04Nuclear boiler instrumentation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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Low Source: NEW-CONTRACT Which of the following design features associated with the plant's Safety Relief Valves (SRVs) functions to mitigate containment loads c aused by reopenings of an SRV by reducing the frequency of subsequent SRV actuations following the initial SRV opening.A.Two-Stage Target Rock ValvesB.SRV discharge line T-quenchers C.Low-Low Set relief feature of two SRVs D.Vacuum-relief valves located in each SRV discharge lineAnswer:C ILO 2015 Written Page: 99 of 218 25 June 2015 Answer Explanation:

The examinee should correctly determine that it is the LLS relief feature of two of the plant's SRVs that serve to allow time for the water leg that forms in the SRV discharge piping following SRV closure (from discharge piping residual steam condensation) to clear. The examinee should conclude that eliminating the water leg reduces the loading from subs equent SRV actuations to acceptable levels.

Distracter Explanation:A.Is incorrect but plausible because the examinee could incorrectly determi ne that use of two-stage Target Rock valves at Fermi 2 was done to allow proper blowdown time in order to allow the water

leg to drain from the SRV discharge piping prio r to a subsequent discharge. The examinee could incorrectly assume that the LLS Relief function and use of Two-stage Target Rock valves work in conjunction to perform this functi on when, in fact, they work toget her to meet the requirements of NUREG-0737 to reduce the frequency of stuck open sa fety relief valve events at Fermi 2. B.Is plausible because the SRV discharge line T-quenc hers are part of the overall SRV design that limits loading forces on containment and the examinee could incorrectly determine that the holes in

the T-quencher are what limits the leg of water in the SRV tailpipe and forget that the purpose of the SRV T-quenchers is to limit valve outlet pressure to 40 percent of maximum valve inlet pressure through the use of the holes drilled in the termination pipe.D.Is plausible because the vacuum relief feature (vacuum breakers) of the SRV discharge lines are part of the overall SRV design that limits forces on containment and the exam inee could incorrectly conclude that the vacuum breakers allow time for t he leg of water to clear without recalling that the vacuum relief valves provided on each SRV dischar ge line prevent drawing an excessive amount of water up into the line as a result of steam condensation following termination of relief operation.

Reference Information:T.S.B 3.6.1.6 Low-Low Set (LLS) Valves (pg B 3.6.1.6-1 to 2)

APPLICABLE SAFETY ANALYSES NUREG 1123 KA Catalog Rev. 2239002 K4.Knowledge of RELIEF/SAFETY VALVES desi gn feature(s) and/or interlocks which provide for the following:239002 K4.03Prevents siphoning of water into SRV discharge piping and limits loads on subsequent actuation of SRV's Technical Specifications3.6.1.6 Low-Low Set (LLS) Valves 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: MODIFIED BANK: 20201420011001 Which of the following indications are correct for SRV operation?A.A & CB.B & C C.A & D D.B & DAnswer:B ILO 2015 Written Page: 101 of 218 25 June 2015 Answer Explanation:

Graph for opening and then closing of a SRV Distracter Explanation:

Distracters are plausible based on Reactor response to SRV open.

Reference Information:

AOP 20.000.25 (pg 7)

ILO 2015 Written Page: 102 of 218 25 June 2015 NUREG 1123 KA Catalog Rev. 2239002 A1.Ability to predict and/or monitor changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including:239002 A1.05Reactor water level 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

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High Source: MODIFIED BANK: 3150146C002003The plant is operating at 50% power. Digital Feedwater Level Control is in 3-Element control with the Reactor Level Select Switch in A.

3D164, FEEDWATER CONTROL DCS TROUBLE, has alarmed, and both Reactor Feedwater Pump Controllers have switched to Emergency Bypass.

(1) What is the resulting impact to Feedwater Level Control (DCS), and (2) what action is required by procedure?A.(1) DCS will shift to Forced Single Element Control.

(2) Continue power operation in Single Element Control per SOP 23.107.B.(1) DCS remains in 3 Element Control.

(2) Manually adjust Reactor Feedwater Pu mp Controllers to match Feed Pump speeds per SOP 23.107.C.(1) DCS will shift to Forced Single Element Control.

(2) Since Adequate Pumping Capacity is NOT available, shutdown the reactor per AOP 20.107.01.D.(1) DCS remains in 3 Element Control.

(2) Manually adjust Reactor Feedwater Pu mp Controllers to match Feed Flow with Steam Flow per AOP 20.107.01.Answer:D ILO 2015 Written Page: 104 of 218 25 June 2015 Answer Explanation:

With both Reactor Feed Pump Controllers in EMERGENCY BYPASS, manual feedwater control is required by the AOP to maintain level.

Distracter Explanation:A.Is plausible; would be true for a loss of one feed flow or two steam flow inputs, if NOT in Emergency Bypass.B.Is a plausible misconception in that matching RFP speeds would be effective vice matching steam flow and feed flow.C.Is plausible misconception; testing knowledge of pumping capacity with c ontrollers in Emergency Bypass Reference Information:

ARP 3D164 (pg 2)

AOP 20.107.01 (pg9)

Plant Procedures 20.107.01 03D164 NUREG 1123 KA Catalog Rev. 2259002 A2.Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM ; and (b) based on those pr edictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:259002 A2.06Loss of controller signal output 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 105 of 218 25 June 2015 ID: R48 Points: 1.00Difficulty:

0.00 Level of Knowledge:

Low Source: BANK: EQOP3150120B003005 During cooldown operation of Div 1 SGTS, 8D48, DIV I SGTS CO2 DISCH VLV OPEN, alarms.

Which of the following is the LOWEST SGTS Charcoal Bed Temperat ure consistent with this indication?A.150°FB.255°F C.310°F D.355°FAnswer:C ILO 2015 Written Page: 106 of 218 25 June 2015 Answer Explanation:

310°F is the setpoint at which the CO2 discharge valve (F413A/B) will open. Once the valve is open the CO2 pressure will be sensed and bring in the alarm 8D48.

Distracter Explanation:A.Is plausible and incorrect because 150°F is normal Charcoal Adsorber Blanket Heater setpointB.Is plausible and incorrect because 255°F is Cooling Fan Auto Operation setpoint D.Is plausible and incorrect because 355°F is in excess of CO2 Initiation setpoint Reference Information:

T46K002A Setpoints from CECO I-2642-05 (F413A/B actuation logic)

Plant Procedures 08D48 NUREG 1123 KA Catalog Rev. 2261000 A3.Ability to monitor automatic operat ions of the STANDBY GAS TREATMENT SYSTEM including:261000 A3.03Valve operation261000 A3.04System temperature.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 107 of 218 25 June 2015 ID: R49 Points: 1.00Difficulty:

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High Source: NEWThe plant is operating at 100% reactor power with EDG 12 currently synced to the GRID and being controlled from the Main Control Room.

A loss of the 345 Kv mat and a Reactor Scram then occurs.

Based on these conditions, which of the following statements correctly describes EDG operation and parameter response when controlled from the Main Control Room?A.When EDG 12 Governor Control switch is taken to raise, EDG 12 Generator frequency will increase.B.When EDG 12 EDG Voltage Control switch is taken to raise, EDG 12 Generator Voltage will increase.C.When EDG 13 EDG Voltage Control switch is taken to raise, EDG 13 Generator KVARS will increase.D.When EDG 13 Governor Control switch is taken to raise, EDG 13 Generator frequency will increase.Answer:D ILO 2015 Written Page: 108 of 218 25 June 2015 Answer Explanation:

To answer this question the examinee must deter mine that EDG 12 would be in Droop mode and EDG 13 would be in isochronous. Understanding th is is critical in controlling the EDG and being able to keep it's parameter in band as an operator. Based on this t he Governor control switch and the Voltage control switch effect the EDG differently. For DROOP, Governor Control controls LOAD in KW and Voltage Control controls reactive load in VARS. For ISOCHRONOUS Governor Control controls frequency in Hz and Voltage Control controls output voltage.

Distracter Explanation:A.is plausible and incorrect because it w ould be true if EDG 12 was in isochronous.B.is plausible and incorrect because it w ould be true if EDG 12 was in isochronous.C.is plausible and incorrect because it would be true if EDG 13 was in droop.

Reference Information:

ST-OP-315-0065-001 (pg 26)

Plant Procedures 23.307 NUREG 1123 KA Catalog Rev. 2262001 A4.Ability to manually operate and/or monitor in the control room:262001 A4.05Voltage, current, power, and frequency on A.C. buses 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: NEW-CONTRACT A fire in the BOP switchgear room has c aused several BOP busses to be de-energized and resulted in the following indications being observed in the Control Room: *3D22, UPS UNIT A/B TROUBLE, alarmed. *11D42, SWYD DCS TROUBLE, alarmed.

All C32, Reactor Feedwater DCS, i ndications on panel H11-P603 have been lost. *RPIS indications on the Full Core Display are de-energized.

The Rod Worth Minimizer is de-energized.

What is the status of the UPS system, and what actions are necessary to restore control of the 120kV mat and CTGs from the Fermi 2 Contro l Room and the H21-P623 Dedicated Shutdown Panel?A.UPS A and B are de-energized; temporar y power needs to be provided to BOP BUS 72L.B.UPS A and B are de-energized; tempor ary power needs to be provided to 120Kv/SBO DCS.C.UPS A is de-energized; U PS A needs to be manually transferred to its Alternate Power Supply using the Bypass to Alt Line switch.D.UPS B is de-energized; U PS B needs to be manually transferred to its Alternate Power Supply using the Bypass to Alt Line switch.Answer:B ILO 2015 Written Page: 110 of 218 25 June 2015 Answer Explanation:

Based on the alarm conditions and indications provi ded in the Main Control Room, UPS A and B is de-energized. Also, power has been lost to UPS C, the 120Kv/SBO DCS unit, and temporary power will need to be provided in order to restor e control to the affected components.

Distracter Explanation:A.Is plausible because the examinee could evaluate the control room indications and recognize that UPS A and B are de-energized and determine incorrectly that the course of action would be to supply temporary power to UPS how ever, 72L will not provide power to the UPS system, the normal power and alternate is 72M and 72R, power to either of which could power UPS A and B.C.Is incorrect but plausible because the examinee c ould incorrectly determine, from the indications given, that only UPS A was de-energized.

If only UPS A was de-energized, then manually transferring UPS A to its alternate source using the Bypass to Alt Line switch could be a viable option since the UPS B source would still be av ailable as the alternate to UPS A.D.Is incorrect but plausible because the examinee c ould incorrectly determine, from the indications given, that only UPS B was de-energized.

If only UPS B was de-energized, then manually transferring UPS B to its alternate source using the Bypass to Alt Line switch could be a viable option

since the UPS A source would still be av ailable as the alternate to UPS B.

Reference Information:ARP 3D22 Auto action show UPS A -> ADS Timer.

SOP 23.308.01 P&L 3.10 on pg 6 Plant Procedures 03D022 23.308.01 NUREG 1123 KA Catalog Rev. 2G2.2.44Ability to interpret control room indi cations to verify the status and operation of system, and understand how operator actions and directives affect plant and system conditions262002UPS (AC/DC) 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

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High Source: NEW The BOP D.C. Ground Detection Panel has the fo llowing indications when the left pushbutton is depressed. Based on these indications, use 23.

309 Enclosure A, to identify the fault. (See attached 23.309 Enclosure A.)A.NoneB.Fault on positive wiring C.Fault on neutral wiring D.Fault on negative wiringAnswer:C ILO 2015 Written Page: 112 of 218 25 June 2015 Answer Explanation:

Answer is based on 23.309 Encl A. The light bri ghtness that is displayed for this question is (1) Dim (2) Dim (3) Out (4) Very Dim (5) Very Dim (6) Very Dim

With PB #1 being in "B" and PB # 2 being in "A" Distracter Explanation:

Distracters are plausible based on understanding t he indications and correctly using Encl A.

Reference Information:

23.309 Encl A Plant Procedures 23.309 NUREG 1123 KA Catalog Rev. 2263000 K1.Knowledge of the physical connections and/or cause effect relationships between D.C. ELECTRICAL DISTRIBUTION and the following:263000 K1.04Ground detection 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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3.00 Level of Knowledge:

High Source: BANK: 315-0171-A014-001(M)

The plant is operating at 100% power wi th the following auxiliary equipment lineup: *East Station Air Compressor running; Center in Auto *South H2 Seal Oil Pump running; North in Auto

North RFPT West Lube Oil Pump running; East in Auto

North and Center TBCCW pumps running.

Bus 72N is lost due to an internal electrical f ault. What is the appropria te operator response to this event?A.Perform a rapid power reduction.

B.Start both SBFW pumps and inject at 1200 gpm.

C.Verify the South TBCCW pump has automatically started.

D.Verify the Center Station Air Co mpressor has automatically started.Answer:D ILO 2015 Written Page: 114 of 218 25 June 2015 Answer Explanation:

The East Station Air Compressor is powered from 72N. The operator must verify the Center standby equipment auto starts per 20.300.72A Condition C.

Distracter Explanation:A.is incorrect but plausible because the examinee could incorrectly determine that this action is required based on only a single TBCCW pump being available. B.is incorrect but plausible because the examinee could incorrectly determine the loss of the North RFPT Lube oil pump and possible North RFPT trip. C.is incorrect because the South and Center TBCCW pump power is from 72N and TBCCW pumps do not auto start.

Reference Information:

AOP 20.300.72A (pg 3) Condition C Plant Procedures 20.300.72N NUREG 1123 KA Catalog Rev. 2300000 K2.Knowledge of electrical power supplies to the following:300000 K2.01Instrument air compressor 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam 2012 ILO Audit Exam ILO 2015 Written Page: 115 of 218 25 June 2015 ID: R53 Points: 1.00Difficulty:

2.00 Level of Knowledge:

Low Source: BANK: 315-0067-C005-001 With the plant operating at full power, ALL Reactor Building Closed Cooling Water (RBCCW)

AND Emergency Equipment Cooling Water (EECW) flow is LOST. Which ONE of the following components requires action to be taken within TWO minutes, without regard to temperature change?A.HPCI Pump Room CoolerB.Reactor Recirculation Pumps C.Control Rod Drive Hydraulic Pump D.Reactor Building Steam Tunnel Space CoolersAnswer:B ILO 2015 Written Page: 116 of 218 25 June 2015 Answer Explanation:

Reactor recirculation pump operation is limited to TWO minutes without RBCCW/EECW flow, Overide statement for 20.127.01 Distracter Explanation:

Distractors are valid based on being loads on the system addressed by the AOP.

Reference Information:

AOP 20.127.01 (pg 2)

Plant Procedures 20.127.01 NUREG 1123 KA Catalog Rev. 2400000 K3.Knowledge of the effect that a loss or malfunction of the CCWS will have on the following:400000 K3.01Loads cooled by CCWS 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 117 of 218 25 June 2015 ID: R54 Points: 1.00Difficulty:

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High Source: NEW For a normal NOTCH IN of a selected control rod using RMCS, which of the following plots show how the system parameters will respond?

NOTE - For these plots the ROD MOVEMENT CONTROL SW is taken to NOTCH IN, then released.A.A & CB.A & D C.B & C D.B & DAnswer:A ILO 2015 Written Page: 118 of 218 25 June 2015 Answer Explanation:This question asks the examinee to predict the response of Drive Water Flow and its most related parameter Drive Water Diff Pressure.

A and C plots are the Normal system response.

Distracter Explanation:

B Plot is plausible and incorrect because a NOTCH OUT would provide a flow of 2 gpm

D Plot is plausible and incorrect because it shows t he normal cycle of Drive Water Diff pressure, however it is for a NOTCH OUT.

Reference Information:

SOP 23.623 reference provided for understanding of nor mal system operations. Graphs are plot data based on system operations for NOTCH IN and OUT.

Plant Procedures 23.623 NUREG 1123 KA Catalog Rev. 2201002 A1.Ability to predict and/or monitor changes in parameters associated with operating the REACTOR MANUAL CONTROL SYSTEM controls including:201002 A1.01CRD drive water flow 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

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High Source: NEW-CONTRACT The plant was operating at 100% power with C ontrol Rod 14-47 at Notch Position 20 when the following series of events occurred:*At 10:00 Control Rod 14-47 started drifting due to a stuck collet assembly.

At 10:02 Control Rod 14-47 was verified to be fully inserted.

At 10:05 Control Rod 14-47 was hydraulically disarmed.

At 10:15 Control Rod 14-47 was individually scrammed.

Assuming these were the only operator actions performed, whic h of the following describes how reactor power was affected duri ng the above series of events?A.Power was increasing at 10:00 and again at 10:05.B.Power was decreasing at 10:00 and then increasing at 10:05.

C.Power was decreasing at 10:00 and remained steady after 10:02.

D.Power was increasing at 10:00 and remained steady after 10:02.Answer:A ILO 2015 Written Page: 120 of 218 25 June 2015 Answer Explanation:

A stuck collet assembly would cause power to increas e as the rod drifted out of the core. The rod was inserted by operator Immediate Acti on at 10:02 and, after being disarmed at 10:05 would drift back out of the core again causing power to increase. The examinee should verify this by the fact that the rod was individually scrammed at 10:15, which is only necessa ry if the rod continued to drift out after being disarmed, which is another symptom of a stuck collet assembly.

Distracter Explanation:B.Is plausible because the examinee could incorrect ly determine that a stuck collet assembly would cause control rod 14-47 to drift into the core, which would require the operator to verify the control rod fully inserted and then take action to disarm the c ontrol rod. The operator could then incorrectly determine that the stuck collet would then cause t he rod to drift out, once hydraulically disarmed, which would require further action to individually scram the control rod.C.Is incorrect but plausible because the examinee could incorrectly determine that a stuck collet assembly would cause control rod 14-47 to drift into the core, which would require the operator to verify the control rod fully inserted and then take acti on to disarm the control rod, which would prevent any further power changes.D.Is plausible because the examinee should correctly determine that a stuck collet assembly would cause power to increase as the rod drifted out of the core. Howeve r, the candidate could incorrectly conclude that, once the control rod was inserted at 10:02 and subsequently disarmed at 10:05, these actions would hold control rod 14-47 fully insert ed and that individually scramming the rod was only necessary for procedural compliance.

Reference Information:

AOP 20.106.07+ BASES Caution 1 & Actions D.1-D.6 Plant Procedures 20.106.07 20.106.07 Bases NUREG 1123 KA Catalog Rev. 2201003 K3.Knowledge of the effect that a loss or malfunction of the CONTROL ROD AND DRIVE MECHANISM will have on following:201003 K3.01Reactor power 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: NEW-CONTRACT A reactor startup is in progress with power in the Source Range.During turnover, the oncoming RO reviewed the current and following steps of the Rod Withdrawal Sequence and associated rod positions and identified the following:

Step Rod Move From/To Current Position 27 14-39 08 to 12 12 38-15 08 to 12 12 38-39 08 to 12 12 14-15 08 to 12 10 28 30-31 08 to 12 08 22-31 08 to 12 08 30-23 08 to 12 08 22-23 08 to 12 08 All other Control Rods are at their target posit ions and the Rod Worth Minimizer does not indicate any errors or blocks.

After turnover, the RO turns on Rod Select Po wer and selects Control Rod 30-31 for movement.

Which of the following lists ALL of the errors or blocks that will be displayed on the Rod Worth Minimizer?

Key: SE = Select Error; IB = Insert Block; WB = Withdrawal Block; IE = Insert Error; WE =

Withdrawal ErrorA.ONLY SEB.ONLY IB and WB C.ONLY SE, IB and WB D.SE, IB, WB, IE and WEAnswer:C ILO 2015 Written Page: 122 of 218 25 June 2015 Answer Explanation:

Control Rod 30-31 is not the next rod in the s equence and a SE will be displayed. The SE generates both a WB and an IB and will not permit any rod motion. Wi th all rods still in their target positions, neither an IE nor a WE will be displayed.

Distracter Explanation:A.Is plausible because the examinee could recognize that Control Rod 30-31 is not the next rod in the sequence, so a SE will be displayed; however the examinee could incorrectly determine that the SE is all that is generated and fail to recognize that a SE also generates both a WB and an IB.B.Is incorrect but plausible because the examinee c ould correctly recognize that Control Rod 30-31 is not the next rod in the sequence and determine that both an IB and WB are generated without recognizing that the IB and WB are generated as a re sult of the SE, which will also be displayed.D.Is plausible because the examinee could recognize that Control Rod 30-31 is not the next rod in the sequence, so a SE will be displayed. The exami nee could also determine that the SE generates both a WB and an IB and will not permit any rod motion.

However, the examinee could incorrectly conclude that both an IE and WE are also generated, and will be displayed.

Reference Information:

ST-OP-315-0013 (pg 11) explans Rod Block and Error Functions Plant Procedures 23.608 NUREG 1123 KA Catalog Rev. 2201006 K4.Knowledge of ROD WORTH MINIMI ZER SYSTEM (RWM) (PLANT SPECIFIC) design feature(s) and/or interlo cks which provide for the following:201006 K4.03Select blocks/errors: P-Spec(Not-BWR6) 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: NEW-CONTRACT The plant has experienced a loss of bus 72B. While the crew is performing actions in accordance with 20.300.72B to stabilize the plant, a failure of BOTH seals for B Reactor Recirculation pump occurs.Which of the following valves will the crew NOT be able to close to isolate B RR Pump?A.B3105-F023B, South RR Pump Suction Valve B.B3105-F031B, South RR Pump Discharge Valve C.G3352-F106, RWCU RR Loop B Suction Isolation Valve D.B3100-F008B, South RR Pump Seal Water Isolation ValveAnswer:A ILO 2015 Written Page: 124 of 218 25 June 2015 Answer Explanation:

B3105-F023B needs to be closed in order to isolat e B RR pump, but is powered by bus 72B, and will NOT be able to be closed. This question is challenging because BOTH of the RR Pumps' suction isolation valves (F023A & B) are powered by the same bus, 72B, which is a Div 1 bus, and RR Pump B

has primarily Div 2 powered components.

Distracter Explanation:B.Is plausible because B3105-F031B needs to be cl osed in order to isolate B RR pump, and the examinee could fail to recognize that this valve is still powered from bus 72CF so it IS able to be closed to isolate B RR Pump.C.Is plausible because G3352-F106 needs to be cl osed in order to isolate B RR pump, and the examinee could fail to recognize that this valve is still powered from MCC 72E so it IS able to be closed to isolate B RR Pump.D.Is plausible because B3100-F008B needs to be cl osed in order to isolate B RR pump, and the examinee could fail to recognize that this valve is normally directed to be closed locally, so it IS able

to be closed to isolate B RR Pump.

Reference Information:

20.300.72N Encl A pg 3 of 4 LIST OF AFFECTED LOADS Plant Procedures 23.138.01 20.300.72B NUREG 1123 KA Catalog Rev. 2202001 K2.Knowledge of electrical power supplies to the following:202001 K2.03Recirculation system valves 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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5.00 Level of Knowledge:

High Source: MODIFIED BANK:20204010203009 The plant is in Mode 2 with reactor heatup in progress. The Reactor Water Cleanup System (RWCU) is lined up to blowdown from RWCU to the Main Condenser. The following events then occur: 2D119, RBCCW PUMPS DIFF PRESS HIGH / LOW, alarms.2D46, MOTOR TRIPPED, alarms. Both operating RBCCW pumps indicate TRIPPED. NO operator actions have been taken.

(1) For RWCU what automatic actuations will occur?

(2) How will RWCU respond to these actuations?A.(1) G3352-F119, RWCU Inlet Isolation Valve, closes.

(2) RWCU pumps trip on low flow.B.(1) G3352-F044, Filter/Demineralizer Bypass Valve, opens.

(2) RWCU Filter/Demineralizers outlet temperature will stablize and begin to lower.C.(1) G3352-F220, G3352-F004, and G3352-F001, RWCU Containment Isolation Valves, close.

(2) RWCU pumps trip on low flow.D.(1) G3300-F033, Blowdown Flow C ontrol Valve, throttles closes.

(2) RWCU Filter/Demineralizers outlet temperature will stablize and begin to lower.Answer:A ILO 2015 Written Page: 126 of 218 25 June 2015 Answer Explanation:

Loss of RBCCW cooling to NRHXR causes a high temperature which leads to this effect.

RWCU NRHXs NRHX Outlet Temp at 140°F:

G3352-F119 closes.

RWCU Pumps trip.

RWCU Demins into Hold.

Distracter Explanation:

All distractors are plausible and are based on the exam inees ability to properly interpret indications and understand system operations.

Reference Information:

ARP 2D110 (pg 1) AUTO ACTION Plant Procedures 20.127.01 02D110 NUREG 1123 KA Catalog Rev. 2204000 A3.Ability to monitor automatic operat ions of the REACTOR WATER CLEANUP SYSTEM including:204000 A3.03Response to system isolations 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: NEW-CONTRACT The plant is operating at 100% power. A fire in the Reactor Building is reported affecting both Reactor Pressure Vessel (RPV) Level Instrument racks.

If only the Wide Range RPV level instruments' re ference leg temperatures become elevated due to the fire, which one of the following correctly completes the following statement indicating the effect on the RPV level actuations as compar ed to the RPV level trip setpoint under normal temperature conditions?

Due to the elevated temperature in t he level instrument reference legs, actual RPV level for a __(1)__ would be __(2)__ when the actuation occurred.A.(1) reactor scram (2) higherB.(1) reactor scram (2) lowerC.(1) core spray logic actuation (2) higherD.(1) core spray logic actuation (2) lowerAnswer:D ILO 2015 Written Page: 128 of 218 25 June 2015 Answer Explanation:

Heating of the reference legs of any RPV level inst rument would cause the indicated level to increase due to the density change of the water in the reference leg. Based on the lowering density in the reference leg, the reference leg would have less mass as compar ed to the variable leg (actual level) thus making actual RPV level lower for any set point initiated actuation or trip.

The wide range instruments also prov ide core spray (ECCS) actuations.

Distracter Explanation:A.Is incorrect but plausible because the exam inee could incorrectly determine that wide range instruments provide reactor scram functions. The reactor scram functions are provided by narrow instruments. The instrument malfunction due to the elevated temperatures is indicated only if the variable leg temperatures were affected and not the reference leg which would be an incorrect assessment.B.Is incorrect but plausible because the exam inee could incorrectly determine that wide range instruments provide reactor scram functions.

The instrument malfunction due to the elevated temperatures is commensurate with the elev ated reference leg temperature and would be an accurate assessment of the effect.C.Is incorrect but plausible because the exami nee could incorrectly det ermine the instrument malfunction due to the elevated temperatures is indi cated only if the variable leg temperatures were affected and not the reference leg which would be an incorrect assessment.

Reference Information:

BC07Sr4_Sensors May 2011 Explains te mperate variations on instruments.

23.601 (pg 16) Core Spray actuation from these instruments / logicI2rprod-CECO - Identifies the instruments listed in 23.601 as the wide range instruments.

Plant Procedures 23.601 29.ESP.01 NUREG 1123 KA Catalog Rev. 2216000 K5.Knowledge of the operational implications of the following concepts as they apply to NUCLEAR BOILER INSTRUMENTATION:216000 K5.14Density 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 129 of 218 25 June 2015 ID: R60 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW The plant is shutdown. A coolant leak re sulted in an automatic scram and an emergency depressurization.

Current conditions are as follows:

Drywell pressure is 19.5 psig. *Torus pressure is 19 psig.

RPV level is 198 inches and steady.

Division 1 & 2 CS are injecting and being used to control RWL.

All RHR pumps are off.

The CRS directs E1150-F010 closed and Torus C ooling and Spray placed in service using Division 1 RHR. Drywell Spray is then placed in service using Division 1 RHR adding an additional 12,500 gpm.

A leak from the Torus then occurs. Initial Torus level is -14 inches and lowering at 2 inches per minute.

How long before the RHR pump(s) providing Torus Cooling and Spray and Drywell Spray would have to be shut off or flow r educed? (See attached 29.100.01 Sheet 6.)A.13 minutesB.18 minutes C.28 minutes D.37 minutesAnswer:B ILO 2015 Written Page: 130 of 218 25 June 2015 Answer Explanation:Div 1 RHR flow for the above condition is ~22859 gpm (2 pump configuration required per SOP)

-14 inches starting level with - 2 inches per minute

Distracter Explanation:

Distractors are valid based on understanding system configurations and expected flows for RHR 12500 gpm ~ -40 inches - 1 pumps (-14 to -40) 26 inches = 13 minutes 23000 gpm ~ -50 inches - 2 pumps (-14 to -50) 36 inches = 18 minutes - CORRECT ANSWER 18500 gpm ~ -70 inches - 2 pumps (-14 to -70) 56 inches = 28 minutes

12500 gpm ~ -88 inches - 2 pumps (-14 to -88) 74 inches = 37 minutes Reference Information:SOP 23.205 Encl A - Procedure for RHR CONTAINMENT COOLING MODES OPERATION (Shows 2 RHR pumps would be used)

EOP 29.100.01 SH6 - RHR (LPCI) Vortex limit Plant Procedures 23.205 29.100.01 SH 6 NUREG 1123 KA Catalog Rev. 2230000 K1.Knowledge of the physical connections and/or cause effect relationships between RHR/LPCI:TORUS/SUPPRESSION POOL SPRAY MODE and the following:230000 K1.01Suppression pool 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 131 of 218 25 June 2015 ID: R61 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW-CONTRACT A plant shutdown is in progress with reactor pow er currently stable at 65%. The Turbine Flow Limiter setpoint is then slowly lowered to 60%.Which one of the following describes the Governor/Pressure Regulator system valve response?A.Turbine Control Valve and Turbine Bypass Valve positions remain the same.

B.Turbine Control Valves throttle close, and Turbine Bypass Valves throttle open.

C.Turbine Control Valve positions rema in the same, and Turbine Bypass Valves throttle open.D.Turbine Control Valves throttle close, and Turbine Bypass Valve positions remain the same.Answer:B ILO 2015 Written Page: 132 of 218 25 June 2015 Answer Explanation:

The setpoint adjustment should have been to maintain the Turbine Flow Limiter 5% above reactor power and not 5% below reactor power. Lowering the turbi ne flow limiter setpoint below the reactor power setpoint will cause the Turbine Control Valves to close. The Turbine Bypass valves will open in response to a reactor pressure increase.

Distracter Explanation:A.Is incorrect but plausible. The examinee could incorrectly determine that setpoint was properly adjusted, which would cause no valve movement since the setpoint difference from reactor power is 5% in the stem. In accordance with 22.000.03, Powe r Operation 25% to 100% to 25%, the setpoint is maintained 5% above (vice below) r eactor power during the shutdown. C.Is incorrect but plausible. The examinee could in correctly determine that t he system response is a combination of responses from distractors A and C which would each be partly correct for the adjustment of the Turbine Flow Limiter and Reactor Flow Limiter.D.Is incorrect but plausible. The examinee could in correctly determine that t he system response would be as stated in distractor A since this would be a correct response if the Reactor Flow Limit setpoint

had been adjusted to 60% instead of t he Turbine Flow Limit setpoint.

Reference Information:

22.000.02 pg 50

22.000.03 pg 8

23.109 pg 38 Plant Procedures 23.109 22.000.02 22.000.03 NUREG 1123 KA Catalog Rev. 2241000 K6Knowledge of the effect that a loss or malfunction of the following will have on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM :241000 K6.12Control/governor valves NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 133 of 218 25 June 2015 ID: R62 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW A reactor operator is performing 24.000.02 Attach ment 1 Eight Hour -- MODE 1,2,3 -- Control Room.The Drywell Floor Drain and Equipment Drain Sump Pumps were placed in RUN, allowed to trip, and returned to AUTO.

The following data is recorded on Data Sheet 1:

TIME *Current Time (hr:min) 0000 *Previous Time (hr:min) 1600 SUMP LEVEL *Floor Drain Sump Level (in.)

26.3 DRYWELL FLOOR DRAIN *Previous Integrator (gal) 119043 DRYWELL EQUIPMENT DRAIN *Previous Integrator (gal) 715937 CURRENT INTEGRATOR READINGS

Assuming any calculated leakage would remain constant for the next 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , would an LCO entry be required and why?A.Yes, due to total leakage.B.Yes, due to unidentified leakage.

C.Yes, due to unidentified leakage and total leakage.

D.No, leakage is within Tech Spec limits.Answer:B ILO 2015 Written Page: 134 of 218 25 June 2015 ILO 2015 Written Page: 135 of 218 25 June 2015 Answer Explanation:LCO 3.4.4RCS operational LEAKAGE shall be limited to:a.No pressure boundary LEAKAGE:

b.< 5 gpm unidentified LEAKAGE:c.< 25 gpm total LEAKAGE averaged over the previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period; andd.< 2 gpm increase in unidentified LEAKAGE previous 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period in MODE 1.

unidentified LEAKAGE --> DRYWELL FLOOR DRAIN.

121539 - 119043 = 2496 gal / 480 min = 5.2 gal/ min ==> LCO ENTRY

identified LEAKAGE --> DRYWELL EQUIPMENT DRAIN.

176701 - 715937 = 764 gal / 480 min = 1.59 gal/min + 5.2 gal/min = 6.79 ==> NOT LCO ENTRY

Distracter Explanation:

Distracters are plausible based on knowledge of T.S. operation leakage and correct monitoring of integrators

Reference Information:

ARP 2D75 DRYWELL SUMP RATE HIGH 24.000.02 Encl filled out 1-31-14 (historical - for data)

24.000.02 Attachment 1 (pg 3-6) the surveillance

T.S. 3.4.4 (pg 3..4-9 to 10) LCO requirements Plant Procedures 24.000.02 NUREG 1123 KA Catalog Rev. 2268000 A4.Ability to manually operate and/or monitor in the control room:268000 A4.01Sump integrators Technical Specifications 3.4.4 RCS Operational LEAKAGE 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 136 of 218 25 June 2015 ID: R63 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: BANK:315-0135-A021-003 With the plant operating at full power, t he following alarms and indications exist:*6D21, E/W OFF GAS RECOMBINER TEMPERATURE HIGH/LOW, alarms.*The West Off Gas Recombiner is in service and is indicating 700°F on N62-R815, Off Gas Components Temperature Recorder.*The East Off Gas Recombiner is in standby and is indicating 270°F on N62-R815, Off Gas Components Temperature Recorder.Which ONE of the following operator actions should be performed to control Off Gas Recombiner Temperature?A.VERIFY N62-F400, 18" Manifold Steam Supply TCV, is OPEN.B.VERIFY N62-F400, 18" Manifold Steam Supply TCV, is SHUT.C.VERIFY N62-N013A, C (West Recombiner Thermostatic Controlled Electric Heaters) at 600°F.D.VERIFY N62-N013D E, F (East Recombi ner Thermostatic Controlled Electric Heaters) at 600°F.Answer:D ILO 2015 Written Page: 137 of 218 25 June 2015 Answer Explanation:

Based on understand the automatic operation of the off gas system the examinee should recognize that the temperature of the standby (East) off gas recombi ner temperature is low (setpoint of 276°F) Based on the ARP, the first action after verifiying the temper ature (provided by the question stem) is to verify the heater setpoint at 600°F.

Distracter Explanation:

A. Is incorrect and plausible because the N62-F400 c ontrols steam flow to the recombiner and by this, temperature. However this is only true for the re combiner in service and the in service recombiner temperature is in band. The examinee would choos e this answer if they incorrecly identified the problem to be in the in service recombiner and wanted to lower temperture.

B. Is incorrect and plausible because the N62-F400 c ontrols steam flow to the recombiner and by this, temperature. However this is only true for the re combiner in service and the in service recombiner temperature is in band. The examinee would choos e this answer if they incorrecly identified the problem to be in the in service reco mbiner and wanted to raise temperture. D. Is incorrect and plausible because this is the Thermostatic Controlled Electric Heaters for the inservice recombine. The examinee would choose this answer if they incorrecly identified the problem to be in the in service recombiner and t hought the heaters could correct the problem.

Reference Information:

ARP 6D21 (pg 1-3) actions and setpoints Plant Procedures 06D21 NUREG 1123 KA Catalog Rev. 2271000 A3Ability to monitor automatic operations of the OFFGAS SYSTEM including:271000 A3.03System temperatures 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 138 of 218 25 June 2015 ID: R64 Points: 1.00Difficulty:

0.00 Level of Knowledge:

Low Source: NEW The CREF System is designed to maintain a habitable environment in the Control Room Envelope for ____ day(s) continuous occupancy a fter a DBA without exceeding Total Effective Dose Equivalent (TEDE) limits.A.1B.7 C.14 D.30Answer:D ILO 2015 Written Page: 139 of 218 25 June 2015 Answer Explanation:

The CREF System is designed to maintain a habitabl e environment in the CRE for a 30 day continuous occupancy after a DBA without exceeding 5 rem Total E ffective Dose Equivalent (TEDE). - from T.S.

Basis B 3.7.3-2.

Distracter Explanation:

1, 7, and 14 day limits are plausible based on there use in tech specs.

Reference Information:

T.S.B 3.7.7 (pg 3.7.3-1 to 2) Limits 3

NUREG 1123 KA Catalog Rev. 2G2.1.27Knowledge of system purpose and or function

.288000Plant Ventilation Systems

.Technical Specifications3.7.3 Control Room Emergency Filtration (CREF) System NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 140 of 218 25 June 2015 ID: R65 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW-CONTRACT Due to radiation monitor alarms, reactor power was lowered in accordance with 20.000.07, Fuel Cladding Failure. The plant is currently operating at 95% power.

The following annunciators and indications are observed:*3D32, DIV I/II RB VENT EXH RADN MONITOR UPSCALE

3D36, DIV I/II RB VENT EXH RADN MONITOR UPSCALE TRIP

White Division 1 Reactor Building Isolate TRIPPED light - ON

White Division 2 Reactor Building Isolate TRIPPED light - OFF Based on these indications, which ONE of the fo llowing describes (1) the status of Secondary Containment and (2) the actions, if any, requir ed by plant procedures assuming all equipment operates as expected.A.(1) Secondary Containment is fully is olated and pressure is being maintained negative.(2) No additional actions are necessary to align required equipment.B.(1) Secondary Containment is NOT fully isolated and pressure is being maintained negative.

(2) Close the open Secondary Containment isolation valves.C.(1) Secondary Containment is fully is olated and pressure is being maintained positive.

(2) Start the non-running division of SGTS to ensure building pressure is lowered to a negative value.D.(1) Secondary Containment is NOT fully isolated and pressure is being maintained positive.

(2) Close the open Secondary Containment isolation valves, and start the non-running division of SGTS to ensure building pressure is lowered to a negative value.Answer:A ILO 2015 Written Page: 141 of 218 25 June 2015 Answer Explanation:

Secondary Containment isolation valves will all be closed even with only one division's trip isolation circuity tripped. When one division trips, isolation valves close in each line for the supply and exhaust.

The other isolation valves will get a close signal w hen the fans trip on low flow. The alarm procedures and AOP 20.000.02, Abnormal Release of Radioactive Materi al, direct verification of actions; however, based on one division's trip circuitry tripping, one divisi on of SGTS would start and is enough to maintain pressure as described in AOP 20.000.02 bases.

Distracter Explanation:B.Is incorrect but plausible because the examinee could incorrectly determi ne that all containment isolation valves are not closed based on only Divi sion 1 secondary containment isolation circuity tripping. C.Is incorrect but plausible because the examinee coul d incorrectly determine that all isolation valves closed and that an additional train of SGTS is required to fully maintain pressure since the system trip circuits would normally start both divisions.D.Is incorrect but plausible because the examinee could incorrectly determi ne that all containment isolation valves are not closed based on only Divi sion 1 secondary containment isolation circuity tripping and an additional train of SGTS is required to fully maintain pressure since the system trip circuits would normally start both divisions.

Reference Information:

AOP 20.000.02 B.1-B.5 + BASES Plant Procedures 20.000.02 23.404 23.426 20.000.07 20.000.02 Bases NUREG 1123 KA Catalog Rev. 2290001 A2.Ability to (a) predict the impacts of the following on the SECONDARY CONTAINMENT ; and (b) based on those predi ctions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:290001 A2.04High airborne radiation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 142 of 218 25 June 2015 ID: R66 Points: 1.00Difficulty:

0.00 Level of Knowledge:

Low Source: NEW 10 CFR 55.25 states "If, during the term of the license, the licensee develops a permanent physical or mental condition that causes the lic ensee to fail to meet the requirements of § 55.21 of this part, the facility licens ee shall notify the Commission..." To ensure that Fermi 2 meets these requirem ents, MGA13, Fermi Medical Requirements, requires that licensed individuals shall be re sponsible to immediately notify _______________ of any change in medical status.A.Medical onlyB.their immediate supervisor only C.Medical and their immediate supervisor only D.Medical, their immediate supervisor, and the Supervisor, Operations TrainingAnswer:D ILO 2015 Written Page: 143 of 218 25 June 2015 Answer Explanation:

MGA 13 Section 2.10.1 states "Licensed individuals sha ll be responsible to immediately notify Medical, their immediate supervisor, and the Supervisor, Operations Training of any change in medical status." Distracter Explanation:

A,B,C. Medical, their immediate supervisor, and S upervisor, Operations Trai ning must be notifed by the Licensed individual immediately.

Reference Information:

MGA 13 pg 13 Plant Procedures MGA13 NUREG 1123 KA Catalog Rev. 2G2.1.4Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, œno-solo operation, maintenance of active license status, 10CFR55, etc.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 144 of 218 25 June 2015 ID: R67 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: BANK:315-0128-A013-001Following a MANUAL Reactor Scram from 50% power, the following conditions are observed in IPCS: PID QUAL VALUE UNITS DESCRIPTION N30DX3017 GOOD 0.00 PCT HP Turbine control valve #1 position N30DX3018 GOOD 0.00 PCT HP Turbine control valve #2 position N30DX3019 GOOD 0.00 PCT HP Turbine control valve #3 position N30DX3020 GOOD 0.00 PCT HP Turbine control valve #4 position S20DC0315 GOOD CLOSED Generator Breaker CM S20DC0315 GOOD CLOSED Generator Breaker CF S13DJ1212 LOW 0.0 MWE Generator Gross Generation The Main Generator Exciter Field Breaker indicates CLOSED.Based on these conditions, which one of the following Turbine Generator Trip signals will initiate?A.Loss of Field B.Reverse Power C.Generator Differential D.Negative Phase SequenceAnswer:B ILO 2015 Written Page: 145 of 218 25 June 2015 Answer Explanation:

With TCVs SHUT, no steam is being supplied to the Turbine Generator. The Generator is still connected to the grid and will begin motoring. Reverse Powe r will be sensed by this condition and will generate a Turbine Generator Trip when the 67 relay is energized. Breaker position must be looked up in SOER.

Distracter Explanation:A.Is incorrect and plausible because it would be true if the Main Generator Fi eld Breaker OPENED, or excitation was lost.C.Is incorrect and plausible would be true if an electric al fault condition occurred resulting in Differential Current condition.D.Is incorrect and plausible would be true if a phase open occurred in the Generator Stator. Like the Reverse Power trip, the Negative Phase Sequenc e trip is time dependent, which is a common misconception.

Reference Information:

AOP 20.000.21 Bases, page 5 Plant Procedures 20.000.21 Bases NUREG 1123 KA Catalog Rev. 2G2.1.19Ability to use plant computer to evaluate system or component status295005 AA1.Ability to operate and/or monitor the following as they apply to MAIN TURBINE GENERATOR TRIP:295005 AA1.04Main generator controls 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

NRC Exam Usage 2015 ILO Exam 2013 ILO Audit Exam ILO 2015 Written Page: 146 of 218 25 June 2015 ID: R68 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW The plant was operating at 100% power with Jet Pump Total Flow of 87.4 Mlbs/hr when reactor power lowered. The following flow indications were observed:

Based ONLY on the indications above, what event has occurred?

A.Jet Pump Failure on Loop A B.Jet Pump Failure on Loop B C.Uncontrolled Recirc Flow change of 5%

on A Reactor Recirc Pump (lowering)

D.Uncontrolled Recirc Flow change of 5% on B Reactor Recirc Pump (rising)Answer:B ILO 2015 Written Page: 147 of 218 25 June 2015 ILO 2015 Written Page: 148 of 218 25 June 2015 Answer Explanation:

20.138.02 Jet Pump Failure Symptoms:

Unexplained change in indicated Core Flow Unexplained change in Recirc Loop Flow Unexplained decrease in Core D/P Jet Pump Percent Differential Pressure deviates ex cessively from the average of the remaining Jet Pump Percent Differential Pressures Additionally the difference between B21-R609B and B21-R609D shows that B loop has a imbalance showing that the jet pump in B loop have a failure.

Distracter Explanation:A. This answer is incorrect because the difference between B21-R609B and B21-R609D shows that B loop has a imbalance showing that the jet pump in B loop.

C. The change in Jet Pump total flow does not ma tch a 5% lowering in A Recirc pump and difference between B21-R609B and B21-R609D shows that B loop has a imbalance showing that the jet pump in B loop has failed. However the flow indications for A loop are lower than B implying (incorrectly) a lowering of A recirc flow.

D. The change in Jet Pump total flow does not matc h a 5% increase in B Recirc pump and difference between B21-R609B and B21-R609D shows that B loop has a imbalance showing that the jet pump in B loop has failed. However the flow indications fo r B loop are higher than A implying (incorrectly) a increase of B recirc flow.

Reference Information:

AOP 20.138.02 Jet Pump Failure (pg

5) Jet Pump Failure Symptoms Plant Procedures 23.138.02 NUREG 1123 KA Catalog Rev. 2G2.1.7Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behav ior, and instrument interpretation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 149 of 218 25 June 2015 ID: R69 Points: 1.00Difficulty:

4.00 Level of Knowledge:

Low Source: BANK: 804-0001-0007-009 With the plant operating at 80% power, at 0800 on August 28, EDG 11 is discovered INOPERABLE.

Which ONE of the following describes LATEST TI ME that SR 3.8.1.1 "Verify correct breaker alignment and indicated power availability for eac h offsite circuit" can be completed WITHOUT entering into a condition which requires a unit shutdown?A.0815 on August 28 B.0850 on August 28 C.0905 on August 28 D.0915 on August 29Answer:B ILO 2015 Written Page: 150 of 218 25 June 2015 Answer Explanation:

SR 3.8.1.1 is due WITHIN ONE HOUR, 0850 Augus t 28 is the latest time which complies.

Distracter Explanation:A.Is incorrect and plausible, but NOT the latest time (15 minutes for event classification)B.Is incorrect and plausible, SR 3.8.1.1 expir ed at 0900 - NO 1.25 extensions are permitted on initial periodD.Is incorrect and plausible, SR 3.8.1.1 expir ed at 0900 - NO 1.25 extensions are permitted on initial period Reference Information:

T.S. 3.8.1 (pg 3.8-1 to 3.8-9)

NUREG 1123 KA Catalog Rev. 2G2.2.36Ability to analyze the effect of ma intenance activities such as degraded power sources, on the status of lim iting conditions for operations Technical Specifications3.8.1 AC Sources Operating 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 151 of 218 25 June 2015 ID: R70 Points: 1.00Difficulty:

2.50 Level of Knowledge:

Low Source: NEW Per MOP05, Control of Equipment, "Access to Prot ected Areas may be granted to allow activities deemed necessary by the Shift Manager." Which of the following personnel are NOT exempt from Protected Equipment restrictions?A.SecurityB.Plant Manager C.NRC inspectors D.Nuclear OperatorsAnswer:B ILO 2015 Written Page: 152 of 218 25 June 2015 Answer Explanation:

Per MOP 5, Control of Equipment, Security, Operat ions, and NRC inspectors are exempt from Protected Equipment restrictions.

Distracter Explanation:

A/C/D are all exempt; so incorrect and plausible.

Reference Information:

MOP05, pages 30 & 31 Plant Procedures MOP05 - Control Of Equipment NUREG 1123 KA Catalog Rev. 2G2.2.14Knowledge of the process for cont rolling equipment configuration or status NRC Exam Usage 2015 ILO Exam 2013 ILO Audit Exam ILO 2015 Written Page: 153 of 218 25 June 2015 ID: R71 Points: 1.00Difficulty:

2.00 Level of Knowledge:

Low Source: BANK: 508-0001-A013-003 Which ONE of the following describes the annual limit s for Total Effective Dose Equivalent (TEDE) as set forth in (1) 10CFR20 and (2) Fermi 2 Administrative Guidelines for persons with Radiation Training and complete current year records?A.(1) 3 rem/year (2) 0.5 rem/yearB.(1) 3 rem/year(2) 1 rem/yearC.(1) 5 rem/year (2) 0.5 rem/yearD.(1) 5 rem/year(2) 1 rem/yearAnswer:D ILO 2015 Written Page: 154 of 218 25 June 2015 Answer Explanation:

10CFR20 limits TEDE to 5 rems/yr, and Administrative limits MRP03 are 1 rem/yr for persons with current records.Distracter Explanation:A.Is incorrect and plausible because 0.5 rem is for the incomplete records.B.Is incorrect and plausible because 3 rem/

year is not the federal TEDE limit.C.Is incorrect and plausible because 0.5 rem/year is the limit with incomplete current year records.

Reference Information:

MRP03 Encl A (pg 1)

Plant Procedures MRP02 NUREG 1123 KA Catalog Rev. 2 G2.3.4 3.2/3.7 Knowledge of radiation ex posure limits under normal and emergency conditions NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 155 of 218 25 June 2015 ID: R72 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW-CONTRACT The plant is operating at 100% power when a LOCA occurs.

Which of the following sets of alarms, if actuat ed during these plant conditions, would indicate a fuel cladding failure requiring use of Standby Liquid Control for Torus pH control?A.3D8 - DIV I / II OFF GAS RADN MONITOR UPSCALE 3D24 - 2 MINUTE HOLDUP PIPE RADN MONITOR UPSCALE TRIPB.3D43 - DIV I / II CONTM AREA RADN MONITOR TROUBLE 3D83 - MN STM LINE CH A / B / C / D RADN MONITOR UPSCALEC.8D1 - TORUS HARD VENT RADIATION HIGH/FAIL 16D8 - TURBINE BUILDING HIGH RADND.3D44 - EFFLUENT PROCESS RADN MONITOR TROUBLE 3D48 - TURBINE BLDG VENT EXHAUST RADN MONITOR UPSCALE / INOPAnswer:B ILO 2015 Written Page: 156 of 218 25 June 2015 Answer Explanation:

20.000.07, Fuel Cladding Failure, and 3D43, DIV I

/ II CONTM AREA RADN MONITOR TROUBLE, state that these alarms are indicative of a gross fuel cladding failure and require use of SLC for torus pH control as directed by both of these procedures.

Distracter Explanation:A.Is incorrect and plausible because the examinee could incorrectly determi ne that the radiation monitor alarms are indicative of a fuel cladding fa ilure. The procedures indicate these alarms would be indicative of a small cladding failure and not a gross failure requiring use of SLC for torus pH control.C.Is incorrect and plausible because the examinee could incorrectly determi ne that a fuel cladding failure would result in a torus hard vent radiation alarm, however in the current plant conditions the vent path would not be aligned and there would be no flow past the radiation monitor. Turbine building high radiation could also be indicative of a fuel clad failure, however procedure direction for this alarm does not direct use of SLC for torus pH control.D.Is incorrect and plausible because the examinee c ould incorrectly determine that 3D44 may alarm for a fuel element failure since 3D44 monitors 9 proc ess rad monitor setpoints an alarms if any are above required trip points. 3D44 does not direct acti ons for the use of SLC for torus pH control.

Similarly 3D48 would be plausible for a cladding fa ilure if radiation was transporting through the turbine building.

Reference Information:

ARP 3D43 (pg 1&2) Subsequent ac tions and Initiating Devices AOP 20.000.07 & BASES (all) gross fuel actions and indications.

ILO 2015 Written Page: 157 of 218 25 June 2015 Plant Procedures 03D083 03D044 03D048 16D08 20.000.02 03D008 03D024 03D043 08D01 20.000.07 NUREG 1123 KA Catalog Rev. 2G2.3.15Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 158 of 218 25 June 2015 ID: R73 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: BANK 315-0141-0003-011 The plant is shutdown for a refueling outage.*RPV temperature is 140°F.*Division 2 RHR is operating in the Shutdown Cooling mode.

RHR Pump B is running.

The following alarms and indications are then noted:*3D156, REACTOR WATER LEVEL LOW

3D79, REAC VESSEL WATER LEVEL L3 CHANNEL TRIP

RPV water level is 160" and steady on Narrow Range level indicators.

RPV water level is 135" and lowering slowly on Wide Range level indicators.Based on these conditions, which ONE of the following actions is taken first in accordance with plant procedures?A.Enter 29.100.01 Sheet 1, RPV Control, and restore level using RPV Flooding.B.Enter 29.100.01 Sheet 1, RPV Contro l, and restore level using Table 1 systems.C.Enter 20.205.01, Loss of Shutdown C ooling, and restore shutdown cooling using B RHR pump.D.Enter 20.205.01, Loss of Shutdown C ooling, and restore shutdown cooling using Division 1 RHR.Answer:B ILO 2015 Written Page: 159 of 218 25 June 2015 Answer Explanation:

RPV Level 3 condition requires entry into the EOPs. It also results in a loss of Shutdown Cooling, but since EOP entry takes priority, actions in the AOP ar e taken after the EOP actions. EOP Sh 1 states to restore level using Table 1 systems. Since RPV le vel CAN be determined, entry into RPV Flooding is not required.Distracter Explanation:A.Is incorrect since RPV level CAN be determined, and entry into RPV Flooding is not required. It is plausible if the examinee incorrectly assumes t he indication discrepancies as a loss of RPV level indication.C.Is incorrect because AOP entry is not a priority during EOP actions. It is plausible because the AOP actions include restoring SDC using the same RHR loop previously in service.D.Is incorrect because AOP entry is not a priority during EOP actions. It is plausible if the examinee incorrectly assumes Div 2 RHR is isolated, and Div 1 RHR must be used to restore SDC.

Reference Information:

29.100.01 Sheet 1 (entry conditions and actions)

Question Cognitive Level Analysis/Synthesis Plant Procedures 20.205.01 29.100.01 SH 1 NUREG 1123 KA Catalog Rev. 2G2.4.1Knowledge of EOP entry conditions and immediate action steps.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam

Early Review Questions for NRC 2015 ILO ILO 2015 Written Page: 160 of 218 25 June 2015 ID: R74 Points: 1.00Difficulty:

2.00 Level of Knowledge:

Low Source: NEW An ALERT Emergency Action Level has been decla red and is NOT security related. The Technical Support Center and Emergency O perations Facility are NOT activated.

Per EP-290, Emergency Notifications, the __(1)__ telephone is normally used to make an INITIAL notification to the US Nuclear Regulat ory Commission which must be completed no later than __(2)__ after the emergency declaration.A.(1) ENS (Emergency Notification System)

(2) 15 minutesB.(1) ENS (Emergency Notification System)

(2) 60 minutesC.(1) HPN (Health Physics Network)

(2) 15 minutesD.(1) HPN (Health Physics Network)

(2) 60 minutesAnswer:B ILO 2015 Written Page: 161 of 218 25 June 2015 Answer Explanation:Per EP-290, the NRC must be notified using the ENS telephone 60 minutes after classifying an event.

Distracter Explanation:A.is plausible because 15 minutes is the requirem ent for local and state authorizes AND the NRC for security events.C.is plausible because the HPN is another communi cation system used for communications directed from EP-290 and 15 minutes is the requirement for local and state authorizes AND the NRC for security events.D.is plausible because the HPN is another communi cation system used for communications directed from EP-290 Reference Information:

EP-290 Page 7 (NRC Notification requirements)

Plant Procedures EP-290 NUREG 1123 KA Catalog Rev. 2G2.4.43Knowledge of emergency communications systems and techniques.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam Early Review Questions for NRC 2015 ILO ILO 2015 Written Page: 162 of 218 25 June 2015 ID: R75 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: BANK: EQOP3150032C001001 During a Loss of Condenser Vacuum, t he following alarm status exists:

4D108, CONDENSER PRESSURE HIGH ALARMING 4D46, MAIN TURBINE TRIPPED ALARMING 3D86, MN STM LINE ISO VALVE CLOSURE CHANNEL TRIP ALARMING5D46, N/S RFPT EXHAUST PRESS HIGH TRIP/FAULT CLEARThese alarms are CONSISTENT with which ONE of the following Main Condenser Backpressure values?A.2.5 psiaB.5.0 psia C.9.5 psia D.12.5 psiaAnswer:C ILO 2015 Written Page: 163 of 218 25 June 2015 Answer Explanation:

Given these alarms, Main Condenser Ba ckpressure is between 6.8 and 12.2 psia.

Distacter Explanation:

A.Is plausible; would be true with ONLY 4D108 (2.21 psia)

B.Is plausible; would be true with 4D108 and 4D46 (3.68 psia)

D.Is plausible; would be true with 4D108, 4D46, 3D86, and 5D46 alarms. (12.2 psia)

Reference Information:

AOP 20.125.01, Loss of Condenser Vacuum, pg 2 Plant Procedures 20.125.01 05D046 04D046 04D108 NUREG 1123 KA Catalog Rev. 2G2.4.46Ability to verify that the alarms are consistent with the plant conditions 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 164 of 218 25 June 2015 ID: S76 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW-CONTRACT The plant is operating at 100% power. An eart hquake occurs and the following indications are observed: 120Kv BUS 101 Voltage indicates 0 AC volts. 120kv breaker positions GM , GK indicate OPEN. 120kv breaker positions GH and GD indicate CLOSED. 345kv breaker positions BM and DM indicate CLOSED. 345kv breaker positions BT, DF , CM, and CF indicate OPEN. 3D73, Trip Actuators A1/A2 Tripped alarms 3D74, Trip Actuators B1/B2 Tripped alarms. 3D85, Primary Containment High Press Channel Trip alarms. Operators place the mode switch in Shutdown, and enter 29.000.01 Sheet 1 RPV control. RPV level is being maintained in the normal band using 'B' SBFW Pump.

Electrical indications are reported by a control room operator as follows:

EDG 11 - Running with an of output 1000 KW EDG 12 - Running with an of output 990 KW EDG 13 - Running with an of output 0 KW EDG 14 - Running with an of output 0 KW No damage is reported on 345kv or 120kv mats.

Based on these indications, what procedural actions would be directed to restore power to de-energized buses?A.Start CTG 11-1, OPEN disconnect GL, and restore power through the Alternate Feed 13.2kv Peaker Bus 1-2B Pos A6.B.Close breaker 13.2kv Peaker Bus 1-2B Pos A6, and restore power using Trans 64 Alternate Feed.C.Start CTG 11-1, OPEN breaker GD , and restore power through Trans 1.D.Manually load EDG 13 and EDG 14.Answer:C ILO 2015 Written Page: 165 of 218 25 June 2015 Answer Explanation:

The SRO examinee must know that the high drywell pressure channel trip would have also initiated an automatic start of all four emergency diesel generator

s. With an automatic start on high drywell pressure, the EDGs will remain running until the high drywell condition is reset. The EDGs will not load automatically load on to their respective busse s without an existing under voltage signal. The SRO examinee must evaluate the KW loading indications given in the question stem and identify that SST64 has been de-energized from offsite power. The norma l power to the EDG 13 and 14 busses is from SST65. Because EDG 13 and 14 are running unloaded, this is an additional indication that SST65 is still energized and the loss of power is isolated to SST

64. The SRO examinee would then evaluate the 20 series Abnormal Operating Procedures to select the correct procedure based on SYMPTOMs (loss of 120kv) and a selection enclosure included in the rela ted electrical AOPs. 20.300.kv would be a correct choice for a loss of power to EDG buses 11 and 12.

The SRO examinee must know that the preferred restoration path as directed by 20.300.120kv is through Trans 1. There are multiple restoration options based on the mechanism for the loss of power.

Distracter Explanation:A.Is not correct and plausible. The SRO examinee could incorrectly determine that power should be restored using CTG-11-1 using the Alternate path. If disconnect GL is closed (normal position) the procedure does not direct opening it unless there is damage to the mat.B.Is not correct and plausible. The SRO examinee could incorrectly determine that the Bus 11 alternate source could be energized based on the 120kv break er line up and choose to restore this power source which would be preferred if available. The examinee would have to have correct knowledge of

the 120kv system flow path for normal and alternate sources.D.Is not correct and plausible. The SRO exami nee could apply the information for 4D32 generator frequency to select 20.300.GRID as the procedure to restore power which would place EDG 13 and 14 in service on respective busses. Also incorrectly identifying that division 2 power is lost based on the indications may lead the SRO examinee to what to preferentially restore t he vital busses, thereby selecting this distractor.

Reference Information:

AOP 20.300.120 (actions for 120 loss)

SD-2500-01 BUS lineup for condition.

Plant Procedures 20.300.Offsite NUREG 1123 KA Catalog Rev. 2295003 Partial or Complete Loss of A.C. Power295003 AA2.Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER :295003 AA2.01 Cause of partial or complete loss of A.C. power NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 166 of 218 25 June 2015 ILO 2015 Written Page: 167 of 218 25 June 2015 ID: S77 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW The plant is operating at 100% reactor power when the Main Control Room receives a report of a confirmed fire in Div 1 Battery Room. The crew enters 20.000.22, Plant Fi res, musters the Fire Brigade, and takes action IAW 20.000.22 and the appropriate fire plan.

When the fire is out the Fire Brigade leader reports that a re-flash watch is set and explains that the Division 1 Battery has been severely damaged.

Which of the following outlines the MINIMUM actions required to be in compliance with Technical Specifications?(See attached T.S. 3.8.4 & T.S. 3.8.5)A.Complete a plant shutdown within 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months .B.Place the Reactor Mode Switch in Shutdown, and be less than 200°F Average Reactor Coolant Temperature within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .C.Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months restore DIV 1 DC to operable with only the DIV 1 Battery Charger in service IAW 23.310 48/24V DC Electrical System.D.Place the Reactor Mode Switch in Startup/Hot Standby within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months , and place the Reactor Mode Switch in Shutdown within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months , and be less than 200°F Average Reactor Coolant Temperature in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .Answer:A ILO 2015 Written Page: 168 of 218 25 June 2015 Answer Explanation:

To answer this question the examinee must first ident ify a required T.S action, by appling a T.S for a system (G 2.2.40) and then identif y which plant conditions meet the required MODE of operation (G 2.2.35).

To determine TS action required (G 2.2.40):

3.8.4 LCO The Division I and Division II DC elec trical power subsyste ms shall be OPERABLE.

The Basis states the battery is requi red for the subsystem to be OPERABLE.CONDITION B. One DC electrical power subsystem inoperable for reasons other than Condition A.

REQUIRED ACTION Restore DC electrical s ubsystem to OPERABLE status within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

CONDITION C. Required Action and Associated Completion Time not met.

REQUIRED ACTION: Be in MODE 3 in 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

To determine actions required to place t he plant in the required mode (G 2.2.35):

Once the examinee determines the requirement to be in MODE 3, then based on Table 1.1-1, the examinee will choose the conditions that correspond to MODE

3. Answer A is the only choice that places the plant in MODE 3.

Distracter Explanation:B.Is the conditions for MODE 4, which is incorre ct (see above). If the candidate was only considering the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months required action of CONDITION C. this is the only distracter that meets that time requirement.C.This answer assumes you can be operable without t he battery which is incorrect. If the battery was not required this answer would be plausible bas ed on CONDITION B. Condtions that do not change MODED.This answer is LCO 3.0.3, howev er is does not apply because T.S.

3.8.4 can be complied with. This answer is several MODES over time.

Reference Information:

TS & TS BASIS 3.8.4 LCO, COND B&C NUREG 1123 KA Catalog Rev. 2 G2.2.40 3.4/4.7 Ability to apply tec hnical specifications for a system G2.2.35 3.6/4.5 Ability to determine Tec hnical Specification Mode of Operation295004 Partial or Complete Loss of D.C. Power Technical Specifications3.8.4 DC Sources Operating 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (1) Conditions and lim itations in the facility license.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 169 of 218 25 June 2015 ILO 2015 Written Page: 170 of 218 25 June 2015 ID: S78 Points: 1.00Difficulty:

0.00 Level of Knowledge:

Low Source: MODIFIED BANK: 8012001001007 A fire is in progress. Turbine Building Ar ea Temperature has exceeded 200°F, and an automatic Reactor Scram has occurred. All control rods ar e fully inserted, and the Reactor Mode Switch is in SHUTDOWN.

With smoke accumulating in the Main Control Room, evacuation to the Remote Shutdown Panel has been directed. At the Remote Shutdow n Panel, the following conditions are noted: *RPV Pressure is 950 psig. *RPV Water Level is 175 inches.

Assuming that COLD Shutdown is desired, whic h one of the following is (1) the procedurally directed method for conducting a cooldown, and (2) what is the LOWEST RPV Pressure allowable within ONE hour? (See attached 20.000.19 Attachment 1)A.(1) LOWER the pressure setting on pressure controllers in Turbine Control Relay Panel.

(2) The LOWEST RPV Pressure allowable within ONE hour is 400 psig.B.(1) OPEN Safety Relief Valves A or B at the Remote Shutdown Panel.(2) The LOWEST RPV Pressure allowable within ONE hour is 400 psig.C.(1) LOWER the pressure setting on pressure controllers in Turbine Control Relay Panel.

(2) The LOWEST RPV Pressure allowable within ONE hour is 450 psig.D.(1) OPEN Safety Relief Valves A or B at the Remote Shutdown Panel.(2) The LOWEST RPV Pressure allowable within ONE hour is 450 psig.Answer:D ILO 2015 Written Page: 171 of 218 25 June 2015 Answer Explanation:RISING Turbine Building Area Temperatures causing a Reactor Scram indicate MSIVs are CLOSED. SRVs will be operated. 20.000.19 limits cooldown to 90°F/hr. The LOWEST RPV Pressure allowable

within ONE hour is 450 psig.

Distracter Explanation:A.is plausable if the examinee thought the cause of the scram was not clos ed MSIVs and thought they were OPEN and the cooldown limit is 100°F/hr.B.is plausable if the examinee thought the cooldown limit is 100°F/hr.C.is plausable if the examinee thought the cause of the scram was not clos ed MSIVs and thought they were OPEN.

Reference Information:

AOP 20.000.19 Cond K pg 13 and Attachment 1, pg 1 Plant Procedures 20.000.19 NUREG 1123 KA Catalog Rev. 2295016 AA2.Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT :295016 AA2.06Cooldown rate 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 172 of 218 25 June 2015 ID: S79 Points: 1.00Difficulty:

0.00 Level of Knowledge:

Low Source: NEW You are the CRS and the plant is operating at 100% reactor power when 3D85, PRIMARY CONTAINMENT HIGH PRESS CHANNEL TRIP, alarms. Div 1 & 2 PC Pressure Recorders indicate 1.7 psig and slowly increasing. O ne (1) minute has passed since the scram, and no operator actions have been taken.1D80, DIV 1/2 EECW/EESW SYS IN MANUAL OVERRIDE, is in alarm. P4400-F601A, Div 1 EECW RETURN ISO VLV, indicates OPEN. P4400-F601B, Div 2 EECW RETURN ISO VLV, indicates CLOSED. P4400-F603A, Div 1 EECW SUPPLY ISO VLV, indicates OPEN. P4400-F603B, Div 2 EECW SUPPLY ISO VLV, indicates CLOSED. Based on these indications, which ONE of the fo llowing actions will the CRS direct for the EECW system?A.Place P4400-M001, DIV 1 EECW ISO OVERRIDE SW, in NORMAL.B.Place P4400-M001, DIV 1 EECW ISO OVERRIDE SW, in NORMAL, and Depress P4400-M042, DIV 1 EECW ISO RESET SW.C.Place P4400-M004, DIV 2 EECW ISO OVERRIDE SW, in NORMAL.

D.Place P4400-M004, DIV 2 EECW ISO OVERRIDE SW, in NORMAL, and Depress P4400-M049, DIV 2 EECW ISO RESET SW.Answer:A ILO 2015 Written Page: 173 of 218 25 June 2015 Answer Explanation:

1D80 indicates that the DIV 1 OR DIV 2 EECW ISO OVERRIDE SW is in Override. With this switch in override EECW will not initiate on high drywell pressure. So the correct action is to place P4400-M001 DIV 1 EECW ISO OVERRIDE SW in NORMAL to allow division 1 EECW to actuate as required.

Distactor Explanation:B.Is incorrect and plausible because the reset isolat ion pushbutton is not required to allow the system to start automaticity, however t he pushbutton is used in the SOP in relation to the OVERRIDE SW to reset EECW.C.Is incorrect because division 2 EECW is already running, this answer is plausible because the examine could interpreted 2D14 and 2D17 as indica tion of a problem with Division 2 EECW rather than normal starting alarms.D.Is incorrect because division 2 EECW is already running, this answer is plausible because the examine could interpreted 2D14 and 2D17 as indica tion of a problem with Division 2 EECW rather than normal starting alarms.

Reference Information:

ARP 1D80 DIV I/II EECW/EESW SYS IN MANUAL OVERRIDE (caution descibes effect of switch and direction to return to normal)

NUREG 1123 KA Catalog Rev. 2G2.4.16Knowledge of EOP implementation hier archy and coordination with other support procedures or guidelines such as operating procedures, abnormal operating procedures, and severe accident management guidelines295018Partial or Complete Loss of Component Cooling Water 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 174 of 218 25 June 2015 ID: S80 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW-CONTRACT The plant is operating at 100% power with refuel floor activities in progress to complete fuel bundle sipping in accordance with MOP 16, Conduct of Refuel Floor Activities (Non-Outage).

Permission has been granted to use the INTERLOCK OVERRIDE key for the refueling platform in accordance with 23.710, Fuel Handling System, for sipping operations.A short time later the following occurs:

16D1, RB REFUELING AREA FI FTH FLOOR HIGH RADN, alarms. *Panel H11-P816 ARM recorder indicates channels 15 and 17 in alarm (peaked at 10 mr/hr).

3D31, DIV1/2 FP VENT EXH RADN MONITOR UPSCALE, alarms. *RB HVAC supply and exhaust fans are running.

The licensed operator assigned to the refueling activities reports that a fuel bundle had been withdrawn to approximately 3 feet of the s pent fuel pool surface, however it has been subsequently lowered into a storage location.

What procedural actions are required to be di rected to address these indications and reports?A.Alert personnel using the plant area alarm and Hi-com, evacuate the Refuel Floor, and monitor SFP level.B.Verify isolation of RB HVAC, and ve rify secondary containment integrity.C.Confirm isolation of RB HVAC and initiation of SGTS.D.Isolate all systems discharging into the area except systems required by EOPs and damage control.Answer:A ILO 2015 Written Page: 175 of 218 25 June 2015 Answer Explanation:The indications identify that a high radiation has occurred on the fifth floor due to a reduction in shielding from raising a bundle too near the surface of t he spent fuel pool. 16D1 RB REFUELING AREA FIFTH FLOOR HIGH RADN alarms direct evacuating the area and entering 20.710.01 REFUELING FLOOR HIGH RADIATION. ARM 15 and 17 in alarm at 10 mr/hr with a setpoint of 9 mr/hr validate the alarming

condition. 20.710.01 REFUELING FLOOR HIGH RADIAT ION Also directs alerting personnel using the plant area alarm and Hi-com, evacuating the refuel floor, monitoring SFP level Distracter Explanation:B.Is incorrect and plausible because the SRO ex aminee may interpret the ARM alarm setpoints incorrectly along with the 3D31 DIV1/2 FP VENT EXH RADN MONITOR UPSCALE alarms and decide that RB HVAC isolation and initiation of SGTS is appropriate as directed by 20.710.01

REFUELING FLOOR HIGH RADIATION. 3D31 DI V1/2 FP VENT EXH RADN MONITOR UPSCALE alarms at 2 mr/hr and the TRIP occurs at 3 mr/hr and is indicated by another set of alarms which is not in based on the stem conditions given. If this incorrect assumption is determined the procedure directs verifying isolation of RB HVAC, verifying secondary containment integrityC.Is incorrect and plausible because the SRO exam inee may interpret the ARM alarm setpoints and 3D31 DIV1/2 FP VENT EXH RADN MONITOR UPSC ALE alarms incorrectly and misapply the 29.100.01 Sheet 5 Secondary Containment entry c onditions. ARMs 15 and 17 are not listed in the table for rad level entry conditions for sec ondary containment and FP VENT EXH RADN MONITOR entry setpoints are 5 mr/hr. The alarms only indica te 2 mr/hr. Escalation into the EOP without the entry conditions would not be appropriate. If this incorrect assumption is determined the procedure directs confirming isolation of RB HVAC and initiation of SGTS in the overrideD.Is incorrect and plausible because the SRO exam inee may interpret the ARM alarm setpoints and 3D31 DIV1/2 FP VENT EXH RADN MONITOR UPSC ALE alarms incorrectly and misapply the 29.100.01 Sheet 5 Secondary Containment entry c onditions. ARMs 15 and 17 are not listed in the table for rad level entry conditions for sec ondary containment and FP VENT EXH RADN MONITOR entry setpoints are 5 mr/hr. The alarms only indica te 2 mr/hr. Escalation into the EOP without the entry conditions would not be appropriate. If in correctly determined to be above max normal the action to isolate all systems discharging into the area except systems required by EOPs and damage

control would be appropriate.

Reference Information:

ARP 16D1 (pg 1) INITIAL RESPONSE

ARP 20.710.01 (pg 4) CONDITION A ILO 2015 Written Page: 176 of 218 25 June 2015 Plant Procedures 16D01 03D031 20.710.01 29.100.01 SH 5 NUREG 1123 KA Catalog Rev. 2295023 AA2Ability to determine and/or interpret t he following as they apply to REFUELING ACCIDENTS:295023 AA2.01Area radiation levels.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 177 of 218 25 June 2015 ID: S81 Points: 1.00Difficulty:

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High Source: BANK: EQOP2020121A002007 During an accident condition AFTER Emergency RPV Depressurization, the following conditions exist:*RPV Pressure is 50 psig.*Drywell Temperature is 250°F, RISING.

Drywell Pressure is 42 psig, RISING.

Torus Pressure is 42.5 psig, RISING.

Primary Containment Water Level is 580 ft, RISING.Which ONE of the following actions is REQUIRED? (See attached 29.100.01 Sheet 6.)A.INITIATE Drywell Sprays per 29.100.01 Sheet 2, PRIMARY CONTAINMENT CONTROL.B.VENT the Drywell IRRESPECTIVE of offsite radioactivity release limits per 29.ESP.07 Section 3.0.C.VENT the Torus REMAINING WITHIN offs ite radioactivity release limits per 29.ESP.07 Section 2.0.D.VENT the Torus IRRESPECTIVE of offsite radioactivity release limits per 29.ESP.07 Section 2.0.Answer:B ILO 2015 Written Page: 178 of 218 25 June 2015 Answer Explanation:

With Drywell Pressure above PCPL, it is required to vent the Drywell IRRESPECTIVE of offsite release rate limits per 29.ESP.07 Section 3.0. 29.100.01 SH2 PCP-11 Distracter Explanation:A.is plausible; would be true for Torus Wate r Level < +45 inches, which is < 560 feet.C. is plausible; would be true for Torus Water Level < 570 feet.D.is plausible; would be true for Torus Water Level < 570 feet..

Reference Information:

EOP 29.100.01 SH2 PCP-11 & Curve Plant Procedures 29.100.01 SH 6 29.100.01 SH 2

29.ESP.07 NUREG 1123 KA Catalog Rev. 2295024 EA2.Ability to determine and/or interpret t he following as they apply to HIGH DRYWELL PRESSURE:295024 EA2.01Drywell pressure.

10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 179 of 218 25 June 2015 ID: S82 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW An ATWS has occurred and the Operating Shift is executing the EOPs. SLC Pump A was started at 12:15 with a SLC tank level of 72 inches.

The P603 operator observes that RWCU Isolates as required and SLC Tank level is trending down normally.

Assuming that the SLC system c ontinues operating normally at time 12:45 which of the following EOP actions would be directed?A.Shutdown SLC pumps B.Depressurize the RPV at < 90°F/hr C.Keep RPV water level between -25 inches and 114 inches D.Restore and keep RPV water level between 173 inches and 214 inchesAnswer:D ILO 2015 Written Page: 180 of 218 25 June 2015 Answer Explanation:

SLC TANK is a 9 feet I.D. x 12 feet high ve rtical cylinder tank. Per DBD (4.2.1 pg 16) 1ft³ = 7.48052 US gal V(tank) = r²h => 3.1415(4.5 ft)²(1ft) = 63.61 ft³ = 475.83 gal/ft ~ 40 Gal per inch in SLC TANK The minimum design flow rate occurs at 1,215 psig and is 41.2 gpm for a SLC pump Per the DBD ( 4.2.3 on pg 17)

<45 inches SLC TANK LEVEL is HOT SHUTDOWN BORON WEIGHT per the EOP 29.100.01 SH 1A Table 15 72 INCHES STARTING LEVEL for the SLC TANK 30 Minutes at 42.2 gpm = 1236 gal

At 40 Gal per inch the level drop is 30.9 Inches

41.1 INCHES <-- Current Level EOP FSL-OR2 directs restoring normal water level 173-214 inches on HOT SHUTDOWN BORON WEIGHT Distracter Explanation:A.EOP 29.100.01 SH 1A STEP FSQ-19 directs this when SLC Tank is empty, if the examinee calculates the tank as empty they would choose this answerB.EOP 29.100.01 SH 1A STEP FSP-5 directs cooldown, however this is only with no boron injection or COLD S/D born weight, this answer is plausible if the examinee believes cold S/D born weight has been injected.C.EOP 29.100.01 SH 1A STEP FSQ-OR1 provides for this however, the conditions are not met. The examinee would choose this answer if they assum ed that Hot Shutdown Boron Weight or that Rx power <3% with boron injection allows for the exit of the power leg of SH 1A.

Reference Information:

EOP 29.100.01 SH 1A (yellow box - TABLE 15)

DBD C41 SLC (pg 16 & 17)

Plant Procedures 29.100.01 SH 1A NUREG 1123 KA Catalog Rev. 2 211000 K5.06 3/3.2 Tank level measurement 295037 EA2.03 4.3*/4.4* SBLC tank level 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 181 of 218 25 June 2015 ID: S83 Points: 1.00Difficulty:

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High Source: BANK: EQOP8023002A003001 While mitigating an ATWS per 29.100.01 Sheet 1A, what is the significance of Torus Water Temperature reaching 115°F while Reactor Po wer is 10%? (See attached 29.100.01 Sheet 6.)A.If Emergency Depressurization is conduct ed at this point, the Heat Capacity Limit will NOT be exceeded.B.If Torus Water Temperature continues to increase AND is being used as the injection source, Reactor Power will LOWER.C.If Standby Liquid Control is injected at this point, Hot Shutdown Boron Weight will be injected before the Heat Capacity Limit is reached.D.If ALL injection to the RPV is Term inated and Prevented at this point, RPV Water Level will remain ABOVE TAF when Reactor Power reaches 3%.Answer:C ILO 2015 Written Page: 182 of 218 25 June 2015 Answer Explanation:

BIIT Curve is shown - The sloped part is based on inje cting HSBW prior to exceeding the HCL. The flat part is based on Technical Specifications scram requirement.

Distracter Explanation:

A is plausible; exceeding HCL requires Emergency Depressurization.

B is plausible; graph shows LOWER values of power with HIGHER values of Torus Temperature.

D is plausible; Reactor Power and Torus Temperat ure are considered with RPV Water Level by 29.100.01 Sheet 1A ATWS RPV Control FSL-8, and under these conditions will require Terminating and Preventing Injection.

Reference Information:

BWROG EPG/SAGs Appendix B Bases C5 page B-14-15 Plant Procedures BWROG EPG App B NUREG 1123 KA Catalog Rev. 2G2.1.25Ability to interpret station reference materials such as graphs, curves, tables, etc.295013High Suppression Pool Water Temperature295037SCRAM Condition Present and Reactor Power Above APRM Downscale or unknown.10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 183 of 218 25 June 2015 ID: S84 Points: 1.00Difficulty:

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High Source: BANK: EQOP3150148A018003 The plant is shut down with the following conditions:*Reactor Pressure is 85 psig.*Main Steam Isolation Valves are OPEN.

RHR Loop A is operating in Shutdown Cooling Mode with cooldown in progress.

Reactor Water Cleanup is in service.

Power to RPS Bus A is subsequently lo st due to a trip of RPS MG Set A.

(1) How is the plant affected by this loss, and (2) which of the stated actions will restore the plant to normal?A.(1) B2103-F022 A through D, Inboard Main Steam Isolation Valves, will CLOSE.

(2) Enter SOP 23.137, Nuclear Boiler System. Reopen the MSIVs. Continue cooldown using RHR Loop A.B.(1) G3352-F004, RWCU Supply Outboard Isolation Valve, will CLOSE.

(2) Enter AOP 20.707.01, Loss of RWCU.

RWCU will be restored by placing RPS Div 1 on Alternate Power, resetting the isolation, and realigning RWCU.C.(1) E1150-F017A, Div 1 LPCI Outboard Isolation Valve, and E1150-F008, RHR SDC Outboard Suction Isolation Valve, will CLOSE.

(2) Enter SOP 23.205, RHR System. S hutdown Cooling will be restored by starting an RHR Loop B Pump and opening E1150-F015B, Div 2 LPCI Inboard

Isolation Valve.D.(1) E1150-F015A, Div 1 LPCI Inboard Isolation Valve, AND E1150-F009, RHR SDC Inboard Suction Isolation Valve, will CLOSE.

(2) Enter AOP 20.205.01, Loss of Shutdow n Cooling. Shutdown Cooling will be restored by placing RPS Bus A on Alternate Power, resetting the isolation, and realigning RHR Loop AAnswer:D ILO 2015 Written Page: 184 of 218 25 June 2015 Answer Explanation:RHR Loop A will experience E1150-F015A, Div 1 LPCI Inboard Isolation Valve AND E1150-F009, RHR SDC Inboard Suction Isolation Valve ISOLATION. Wi th RPS on Alternate Power, the isolation may be reset and the system restarted.

Distracter Explanation:A.Is incorrect and plausible; a HALF Isolation c ondition will exist on MSIVs , resulting in no MSIV closures. Actions are correct for an MSIV closure.B.Is incorrect and plausible; tests misconception on logic power, Outboard Valve does NOT ISOLATE, G3352-F001 ISOLATES.C.Is incorrect and plausible; and would be true if E1150 outboards were effected vice inboard isolations.

Reference Information:

SOP 23.601 Pg 11 for E1150-F009 AND E1150-F015A ISOLATION GP AOP 20.205.01 Loss of Shutdown cooling actions.

SOP 23.205.01 Shutdown cooling procedure.

SOP 23.316 Restoring RPS Section 7.3 Plant Procedures 20.205.01 NUREG 1123 KA Catalog Rev. 2295020 AA2Ability to determine and/or interpret the following as they apply to INADVERTENT CONTAINMENT ISOLATION:295020 AA2.06Cause of isolation.

10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 185 of 218 25 June 2015 ID: S85 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW Plant was operating at 100% reactor power when the Reactor Building Vent Exhaust Rad Monitor tripped upscale (3D36). *The CRNSO has reviewed 3D36 and r eports that the alarm is valid. *The source of the release is dischar ging into secondary containment and has not been identified or isolated. *RBSB NE Secondary Containment Rad Lev el has reached Max Norm and is stable. *NO OTHER AREAS are at or approaching Max Safe.

Assuming the plant responded as expected to this alarm, the CRS is required to have entered

__(1)__.

Along with procedural actions that verify the plant responded as expected, the CRS will direct

__(2)__.A.(1)20.000.02, ABNORMAL RELEASE OF RADIOACTIVE MATERIAL only.

(2)Isolate HPCI.B.(1)20.000.02, ABNORMAL RELEASE OF RADIOACTIVE MATERIAL, and29.100.01 SH 5, SECONDARY CONTAINMENT AND RAD RELEASE.(2)Isolate RCIC.C.(1)20.000.02, ABNORMAL RELEASE OF RADIOACTIVE MATERIAL, and29.100.01 SH 5, SECONDARY CONTAINMENT AND RAD RELEASE, and 29.100.01 SH 1, RPV CONTROL.(2)Isolate HPCI.D.(1)20.000.02, ABNORMAL RELEASE OF RADIOACTIVE MATERIAL, and 20.000.21, REACTOR SCRAM, and 29.100.01 SH 5, SECONDARY CONTAINMENT AND RAD RELEASE, and 29.100.01 SH 1, RPV CONTROL.(2)Isolate RCIC.Answer:B ILO 2015 Written Page: 186 of 218 25 June 2015 Answer Explanation:

3D36 in alarm means that RB Vent Exhaust Radiation Monitor is 16,000 cpm increasing. Therefore entry conditions exist for 20.000.02 and 29.100.01 SH 5.

The stem of the questions provides for most of t he required actions, without the source being known.

With 1 max norm the EOP flow charts require isolating all system disc harging into the area therefore the RCIC would need isolated.

Distracter Explanation:

A. is plausible and incorrect because the AOP entry is required and the action is an action the CRS would direct in this case, this answer is incorrect because the CRS is REQUIRED to also enter EOP, additionally the HPCI is not in RBSB NE.

C. is plausible and incorrect if the examinee believes the conditions in the stem will lead to a SCRAM, then entry into the RPV level EOP chart would be r equired and the listed action would be consistent with the RPV level EOP entry, additionally the HPCI is not in RBSB NE. D. is plausible and incorrect if the examinee believes the conditions in the stem will lead to a SCRAM, then entry into the RPV level EOP chart would be r equired and the listed action would be consistent with the RPV level EOP entry and the EOP chart would direct entry into the SCRAM AOP.

Reference Information:

AOP 20.000.02 Plant Procedures 20.000.02 NUREG 1123 KA Catalog Rev. 2G2.4.4Ability to recognize abnormal indications for system operating parameters which are entry-level conditions for emer gency and abnormal operating procedures295034Secondary Containment Ventilation High Radiation 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 187 of 218 25 June 2015 ID: S86 Points: 1.00Difficulty:

3.50 Level of Knowledge:

High Source: BANK: 315-0021-0001-001 RHR is operating in the LPCI mode when Di vision 1 Core Level Instrument fails HIGH.(1) How will automatic RHR System operation be affected, AND (2) what action should be directed as a result?A.(1) LPCI Loop Selection Logic will NOT be able to be reset when RPV Water Level is restored to the desired band.

(2) NO operator action is required at this time.B.(1) Containment Spray Mode Select Interlock will PREVENT operating ONE division of Containment Cooling.

(2) Manually align Containment Cooling isolation valves as neededC.(1) 2/3 Core Height Interlock will NOT operate to automatically isolate Containment Cooling.

(2) Manually isolate that division of C ontainment Cooling if RPV Water Level reaches 2/3 Core Height.D.(1) LPCI initiation on Low RPV Water Level will remain sealed in and will NOT be able to be reset when RPV Water Level is restored.(2) Manually shut down RHR Pumps to terminate LPCI Injection when desired band is reached.Answer:C ILO 2015 Written Page: 188 of 218 25 June 2015 Answer Explanation:

Core Level Instruments only impact the 2/3 Core Cove rage Interlock (-42 inches). One instrument failed high will result in that division not operating to autom atically maximize LPCI Injection. The crew must manually isolate that division of containment cooling if RPV Level reaches -42 inches.

Distracter Explanation:A.Is incorrect because LPCI Loop Select is a wide r ange level function and is plausible if the examinee incorrectly associates LPCI Loop Select.B.Is incorrect because Containment Spray Mode Select is a function of wide range level instrument/High Drywell pressure and will not prev ent operating a division of Containment Cooling.D.Is incorrect because LPCI initiation is a wide range level function is plausible if the examinee incorrectly associates indication with this.

Reference Information:

ST-OP-315-0041-001 was used to develop. Answer Expl anation summarizes system operation that the student text explains. No individual pages are sighted.

See development folder for copy of ST-OP-315-0041-001.Relay progression can be found on I drawings I-2095-03, I-2205-02, and I-2201-71. These show that 2/3 Core Height instruments send a closure signal to contai nment isolation valves to maximize LPCI injection.

If an instrument fails high, that division of containment cooling will not isolate in the event -42 inches is reached, and must be manually isolated.

Instruments B21-N085A/B to trip uni ts B21-N685A/B to relays B21-K217A

/B to B21-K14A/B to contacts for the isolation valves.

ILO 2015 Written Page: 189 of 218 25 June 2015 Plant Procedures 23.205 NUREG 1123 KA Catalog Rev. 2G2.4.21Knowledge of the parameters and logic used to assess the status of safety functions such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, r adioactivity release control, etc.203000 A2. Ability to (a) predict the impacts of the following on the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the cons equences of those abnormal conditions or operations:203000 A2.10Nuclear boiler instrument failures203000 K2.Knowledge of electrical power supplies to the following:

203000 K2.03Initiation logic.

10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam

Early Review Questions for NRC 2015 ILO

2012 ILO Audit Exam ILO 2015 Written Page: 190 of 218 25 June 2015 ID: S87 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW-CONTRACT The plant was operating at 75% power when a trip of the South RFP occurred. Current conditions are as follows:

Recirc pump speeds lower 37% Reactor power 63% SBFW injecting at 600 gpm RPV level 196" and stable HPCI logic then malfunctions causing an automatic initiation of HPCI. Which of the following actions is now required?A.Perform a Rapid Power Reduction.B.Raise North RFP speed C.Inject with SBFW at 1200 gpm.

D.Place the MODE Switch in Shutdown.Answer:D ILO 2015 Written Page: 191 of 218 25 June 2015 Answer Explanation:

HPCI initiation is not within the bound of the analysis for increased core inlet subcooling as described in the AOP bases. Action is to take the MODE Switch to Shutdown.A.Is incorrect because of above and is plausible because it is an Action from 20.107.01 if recirc runback occurs.B.Is incorrect because of above and is plausible because it is an Action from 20.107.01 if inadequate pumping power exists. However, level is stable and manual control of the N RFP is not desired.C.Is incorrect because of above and is plausible bec ause it is an Action from 20.107.01; however, the override takes precedence.

Plant Procedures 20.107.01 NUREG 1123 KA Catalog Rev. 2G2.4.49Ability to perform without reference to procedures those actions that require immediate operation of syst em components and controls.206000HPCI System

.NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 192 of 218 25 June 2015 ID: S88 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW-CONTRACT The following plant conditions exist after a LOCA event:*RPV level

-5 inches*Torus Water Temperature 180°F and stable*Torus Water level

-90 inches and stable*Torus Pressure 4 psig and stable*Reactor Pressure 0 psig and stable*Core Spray Pumps A & C injecting at 6000 gpm*Core Spray Pumps B & D injecting at 3000 gpm Which of the following CS System flows would the CRS direct to maximize CS flow while remaining within operating limits?(See attached 29.100.01 Sheet 6)

A.Lower Div 1 to 5400 gpm.

Raise Div 2 to 8150 gpm.

B.Lower Div 1 to 5400 gpm.

Raise Div 2 to 5400 gpm.

C.Raise Div 1 to 8150 gpm.

Raise Div 2 to 8150 gpm.

D.Raise Div 1 to 7000 gpm.

Raise Div 2 to 7000 gpm.Answer:B ILO 2015 Written Page: 193 of 218 25 June 2015 Answer Explanation:

29.100.01 SH6 Caution 4 states that CS has to be oper ated within NPSH and Vortex limits. These limits are defined by the graphs on SH6.

For a TWL of -90, the MAX flow allowed is 5400 gpm and is the most limiting.

7000 gpm is incorrect because of Vortex Limit and pl ausible if only CS NPSH LIMIT is used with a:

Torus Overpressure = (4 psig) + 3.5 psig + (-90/30)

= 4.5, however SH 6 says do not interpolate, therefore use 0 psig curve for Torus Overpressure. Therefore 180°F TWT = 7000 gpm for 0 psig curve.

8150 gpm is incorrect because of Vortex Limit and plaus ible if only CS NPSH LMIT is used with above torus overpressure and the examinee chooses to interpolate, this gives a value of 8150.

Plant Procedures 29.100.01 SH 6 NUREG 1123 KA Catalog Rev. 2G2.4.20Knowledge of operational implications of EOP warnings, cautions, and notes.209001Low Pressure Core Spray System

.NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 194 of 218 25 June 2015 ID: S89 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW-CONTRACT The plant is in Mode 5 with refueling operations in progress. The refuel floor coordinator reports that a fuel bundle contacted t he edge of a storage rack and was damaged.

Moments later the following indications are observed. *3D31, DIV I / II FP VENT EXH RADN MO NITOR UPSCALE, alarms for all channels. *3D35, DIV I / II FP VENT EXH RADN MONI TOR UPSCALE TRIP, alarms for all channels. *T46-R800A, Div 1 SGTS Flow Recorder, indicates 4950 scfm.

T46-R800B, Div 2 SGTS Flow Recorder, indicates 3950 scfm.

8D46, DIV 1 REACTOR BLDG PRESSURE HIGH/LOW, alarms. *17D46, DIV 2 REACTOR BLDG PRESSURE HIGH/LOW, alarms. *T41-R800A, Div 1 RB Diff Press, indicates -0.6 inches wc.

T41-R800B, Div 2 RB Diff Press, indicates -0.6 inches wc.Based on these indications, (1) determine the impacts for SGTS, and (2) determine what actions are necessary to address the conditions.A.(1) Secondary containment pressure is slightly too negative. The increased flow in Div 1 SGTS increases the charcoal bed adsorption rate.

(2) Maintain Div 1 SGTS in service, and shutdown Div 2 SGTS.B.(1) Secondary containment pressure is slightly too negative. The increased flow in Div 1 SGTS reduces the charcoal bed adsorption rate.

(2) Maintain Div 2 SGTS in service, and shutdown Div 1 SGTS.C.(1) Secondary containment pressure is not negative enough. The decreased flow in Div 2 SGTS increases the charcoal bed adsorption rate.

(2) Maintain Div 2 SGTS in service, and shutdown Div 1 SGTS.D.(1) Secondary containment pressure is not negative enough. The decreased flow in Div 2 SGTS reduces t he charcoal bed adsorption rate.

(2) Maintain Div 1 SGTS in service, and shutdown Div 2 SGTS.Answer:B ILO 2015 Written Page: 195 of 218 25 June 2015 Answer Explanation:

Division one SGTS is operating outside of the nor mal flow range as described in SOP 23.404 as 3879-4180 scfm. The SOP 23.404 will direct stopping one division of SGTS as does ARPs 8D46 and 17D46 based on the secondary containment pressure being slight ly too negative below the alarm setpoint of -0.5 inches wc. Too high of a flow reduces the c harcoal bed adsorption as described in ST-OP-315-0020 page 19.Answer A is incorrect and plausible. The SRO exami nee could incorrectly identify that Division 1 SGTS increased flow would increase the adsorption rate based on more air flow past the bed.

Answer C is incorrect and plausible. The SRO exami nee could identify that Division 2 SGTS flow lower compared would increase the adsorption rate with t he gas remaining near the bed longer. Although this is the normal flow for the system design and does not increase the adsorption rate. Also the answer in incorrect since secondary containment pressure is negative enough.

Answer D is incorrect and plausible. The SRO exami nee could incorrectly identify that Division 2 SGTS flow lower compared would decrease the adsorption rate with the gas remaining near the bed longer.

This is the normal flow range and w ould not decrease the adsorption rate. Also the answer in incorrect since secondary containment pressure is negative enough.

Plant Procedures 17D46 08D46 23.404 NUREG 1123 KA Catalog Rev. 2 261000 Standby Gas Treatment System 261000 A2Ability to (a) predict the impacts of the following on the STANDBY GAS TREATMENT SYSTEM ; and (b) based on those predictions, us e procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:261000 A2.02High system flow NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 196 of 218 25 June 2015 ID: S90 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW The plant is operating at 100% reactor power with the Center station air compressor in service and the East station air compressor in standby.

IAS East DRYER UNIT has been removed from service per 23.129 and tagged out. A malfunction of the IAS West DRYER UNIT starts lowering IAS air pressure. Isolation and bypassing of t he IAS dryer unit is being attempted by field personnel.

This event causes the following plant transient:

IAS HEADER PRESSURE indicates 70 psig and slowly lowering

STATION AIR HEADER PRESSURE indicates 105 psig and steady

NIAS HEADER PRESSURE indicates 97 psig on each division and is cycling within normal band. The following valves have closed:

P5000-F440, DIV 1 CONTROL AIR ISO VLV

P5000-F441, DIV 2 CONTROL AIR ISO VLV

Additionally the following alarms are in:

7D50 DIV I/DIV 2 CONTROL AIR COMPRESSOR AUTO START

7D54 INTERRUPTIBLE CONTROL AIR HEADER PRESSURE LOW

7D56 INTERRUPTIBLE CONTROL AIR DRYER TROUBLE

7D60 RHR COMPLEX CONTROL AIR PRESSURE LOW

3D80 CONTROL ROD DRIFT Which of the following must the CRS direct?A.Place Reactor Mode switch in SHUTDOWN B.Start any available Station Air Compressor.

C.Supply NIAS from Station Air D.Crosstie Div 2 NIAS with IASAnswer:A ILO 2015 Written Page: 197 of 218 25 June 2015 Answer Explanation:

IAS pressure is lowering and 3D80 indicates control r od drift. Therefore the CRS will have to direct the mode switch to shutdown. B.Is plausible and incorrect because it is an action in the AOP for Station Air Pressure < 95. However the IAS air dryer is already passing the maximum cubi c feet of air that it can. The examinee will choose this answer if they believe more air can be pa ssed to IAS from Station air or that Station Air Pressure will lower as IAS lowers so that they m eet the <95 psig to start a Station Air Compressor.C.Is plausible and incorrect because NIAS is normally a load on IAS, however it is not currently a load because the control air compressors are supplying NIAS. The examine will choose this answer if they believe NIAS is still a load for IAS and believe the load reduction can help.D.Is plausible and incorrect because this connection exists in the system via the P500F403, however this connection is only used to s upply NIAS from IAS. There is no procedural that supports for supplying IAS from DIV 2 NIAS because the connec tion is not designed to support suppling IAS because the demand of IAS is much gr eater than DIV 2 NIAS can provide.

Ref 20.129.01 Basis (pg2) and 20.129.01 (override statement)

Plant Procedures 20.129.01 NUREG 1123 KA Catalog Rev. 2300000 A2.Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedur es to correct, control, or mitigate the consequences of those abnormal operation:300000 A2.01Air dryer and filter malfunctions 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 198 of 218 25 June 2015 ID: S91 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW The plant is operating at 100% reactor power. An IAS leak to both CRD flow control valves occurred and IAS is now isolated to both CRD flow control valves.

What is the (1) expected impact and (2) what actions will the CRS order to mitigate this?A.(1) CRD Flow Control Valve Fails CLOSED and Cooling flow to the CRDMs lowers to 15 gpm.

(2) Place the CRD Flow Controller in manual and restore flow.B.(1) CRD Flow Control Valve Fails OPEN and Cooling flow to the CRDMs is maximized.

(2) Restore flow locally using the manual hand jack.C.(1) CRD Flow Contol Valve Fails CLOSED and Cooling flow to the CRDMs lowers to 15 gpm.

(2) Restore flow locally using the manual hand jack.D.(1) CRD Flow Control Valve Fails OPEN and Cooling flow to the CRDMs is maximized.

(2) Place the CRD Flow Controller in manual and restore flow.Answer:C ILO 2015 Written Page: 199 of 218 25 June 2015 Answer Explanation:

Per 20.129.01 Encl B pg 1 C11-F002A(B) CRD Drive Water A(B) FCV

FAIL CLOSE

Cooling flow drops to 15 gpm from 63 gpm (normal se tpoint), because a mechanical block is installed to prevent the flow control valve from fully seating.

20.106.03 provides guidance for manual operation of C11-F002A(B) CRD Drive Water A(B) FCVA. Is plausible and incorrect because the SOP pr ovides guidance for CRD Drive water FCV however local manual however it will use the hand jack not the controller.B. Is plausible and incorrect because the flow cont ol valve fails closed, however the hand jack will be used to open the valve.D. Is plausible and incorrect because the SOP pr ovides guidance for CRD Drive water FCV however local manual however it will use the hand jack not t he controller and the flow contol valve fails closed.Ref (20.129.01Ecnl B pg 1 and 20.106.03 pg7 )

Plant Procedures 20.129.01 NUREG 1123 KA Catalog Rev. 2201001 A2.Ability to (a) predict the impacts of the following on the CRD HYDRAULIC SYSTEM; and (b) based on those predictions, use procedur es to correct, control, or mitigate the consequences of those abnorma l conditions or operations:201001 A2.09Loss of applicable plant air systems.

10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 200 of 218 25 June 2015 ID: S92 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: MODIFIED BANK ID:34016 An ATWS is in progress with the following conditions: *RPV level is being controlled at -15 to 0 inches.

ALL MSIVs are closed.

All Control Rods are NOT inserted.

Torus Cooling, Torus Sprays and Drywell Sprays are being supplied from RHR.

Containment Venting is being performed to maintain Torus Pressure below the Primary Containment Pressure Limit. *The SLC System has failed, and Alter nate Boron Injection (29.ESP.02) is being implemented. *Containment High Range Radiation Monitors indicate 3000 R/hr (40 minutes after the scram).In accordance with EP-101, which ONE of the following Emergency Action Levels is appropriate based solely on Fission Product Barrier Degradation?

(See attached EP-101, Enc. B.)A.Unusual Event FU1 - Any Loss OR Any Potential Loss of Primary Containment.B.Alert FA1 - Any Loss OR Any Potential Loss of EITHER Fuel Clad OR Reactor Coolant System.C.Site Area Emergency FS1 - Loss OR Potential Loss of Any Two Barriers.D.General Emergency FG1 - Loss of Any Two Barriers AND Potential Loss of Third BarrierAnswer:D ILO 2015 Written Page: 201 of 218 25 June 2015 Answer Explanation:

CHRRMS is an indication of Fuel Clad breach, and Venting due to approaching PCPL (due to inadequate Torus Cooling) is RCS and CT Barrier breach.

This question is conceptual on FP Barriers.

Distracter Explanation:A.Is plausible if candidate cannot associate venting irrespective of release rates with CT failure.B.Is plausible if candidate cannot associate CHRRM indication with fuel failure OR Containment Pressure with RCS failure OR associate venting irrespective of release rates with CT failure.C.Is plausible if candidate cannot associate CHRRM indication with fuel failure and Containment Pressure with RCS failure.

Reference Information:

EP-101 TAB F - Provided in Handout folder.

Plant Procedures EP-101 NUREG 1123 KA Catalog Rev. 2G2.4.41Knowledge of the emergency action level thresholds and classifications 219000 RHR/LPCI: Torus/Suppression Pool Cooling Mode NRC Exam Usage 2015 ILO Exam Early Review Questions for NRC 2015 ILO ILO 2015 Written Page: 202 of 218 25 June 2015 ID: S93 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW The plant is operating at 100% power with Div 1 CCHVAC in service when the following occurs:*3D36 DIV I/II RB VENT EXH RADN MONITOR UPSCALE TRIP alarms.*Div I/II RB Vent Exhaust Radiation M onitor reads 16,500 cpm on both divisions.*All automatic actions occur as expected.Based on the above conditions which of the following actions would the CRS direct?A.Verify Group 18 isolationsB.S/D the Rx using GOP 22.000.03 and 22.000.04 C.Shutdown Div 1 CCHVAC Emerg Makeup Fan.

D.Shutdown Div 2 CCHVAC Emerg Makeup Fan.Answer:D ILO 2015 Written Page: 203 of 218 25 June 2015 Answer Explanation:20.000.02 Condition C directs the shutdown CCHVAC Emerg Makeup Fan of the non-operating division if both makeup fans are running.A.Is incorrect and plausable because group 18 is Prim ary Containment Pneumatic Supply System, if the examine does not know the group isolation requirements, this answer is plausible as a penetration of primary containment that could need is olation. Additionally Group 14&16 isolations do occur.B.Is incorrect and plausable because this action is listed in the EOPs however the GOP shutdown is only used if NOTHING is discharging from a prim ary system into secondary containment. This answer is plausible if the examinee feels that t he conditions above merits shutdown, but a scram is undesirable.C.Is incorrect and plausable because a makeup f an does need shutdown, however because Div 1 was in service, Div 2 is the correct answer.

Ref: 20.000.02 pg 3,4,&5 Plant Procedures 20.000.02 29.100.01 SH 5 NUREG 1123 KA Catalog Rev. 2290003 A2.Ability to (a) predict the impacts of the following on the CONTROL ROOM HVAC ;

and (b) based on those predictions, use procedur es to correct, control, or mitigate the consequences of those abnorma l conditions or operations:290003 A2.01Initiation/reconfiguration 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 204 of 218 25 June 2015 ID: S94 Points: 1.00Difficulty:

0.00 Level of Knowledge:

Low Source: NEWYou are the Refuel Floor Supervisor. The plant is shutdown and plant is in MODE 5. Which of the following actions must be completed prior to commencing core alterations?A.Verify 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months have elapsed since Reactor Shutdown.B.Obtain permission to commence core al terations from Operations Engineer.C.Restrict access to the upper elevat ions of the Drywell above the 627 ft.

elevation.D.Confirm the Supervisor, Reactor Engineering shall establish oversight of core alterations from the control room.Answer:C ILO 2015 Written Page: 205 of 218 25 June 2015 Answer Explanation:

MOP 13 requires that prior to Core Alterations that Access to the upper elevations of the Drywell during movement of irradiated core components or fuel will be restricted above the 627 ft. elevation. Access above the 627 ft. elevation but below the 633 ft. elevat ion will be allowed on a case by case basis.

Access is prohibited abov e the 633 ft. elevation.

A. Is incorrect and plauseable because MOP 13 requi res 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months to elaspe since Rx Shutdown before core Alterations. 50 hours5.787037e-4 days 0.0139 hours 8.267196e-5 weeks 1.9025e-5 months is incorrect.

B. Is incorrect and plauseable because MOP 13 requires the Shift Manager give permission for core alterations. The Operations Engineer is the Supervisor for the Shift Managers.

D. Is incorrect and plauseable because the main contro l room does maintain over sight, however this is not what the Supervisor, Reactor Engineering does for core alterations Ref: MOP13 Section 4.1 and Section 4.2 Plant Procedures MOP13 NUREG 1123 KA Catalog Rev. 2G2.1.41Knowledge of the refueling process 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (7) Fuel handling facilities and procedures.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 206 of 218 25 June 2015 ID: S95 Points: 1.00Difficulty:

3.00 Level of Knowledge:

Low Source: BANK ID:27039 During core alterations with fuel movements, in addition to minimum shift complements per MOP03, Policies and Practices, what is the minimum shift complement required per MOP13?

1 Refuel Floor Supervisor, 1 Fuel Handler, _____________A.and no others requiredB.and 1 Fuel Movement Verifier C.1 Fuel Movement Verifier, and 1 Reactor Engineer D.1 Fuel Movement Verifier, 1 Reactor Engineer, and 1 Refuel Floor CoordinatorAnswer:B ILO 2015 Written Page: 207 of 218 25 June 2015 Answer Explanation:

The minimum shift complement for core alterations shall consist of those pos itions listed in MOP03, "Policies and Practices," plus one Refuel Floor Supervisor; and during fuel movements, additional positions requirement of one (1) fuel handl er, and one (1) Fuel Movement Verifier.

A is incorrect because there is no Fuel Movement Verifier C. is incorrect because Reactor Engineer is not required

D. is incorrect because Reactor Engineer and Refuel Floor Coordinator are not required

Reference:

MOP13, page 2 Plant Procedures MOP13 NUREG 1123 KA Catalog Rev. 2G2.1.42 Knowledge of new and spent fuel movement procedures NRC Exam Usage 2015 ILO Exam 2013 ILO Audit Exam ILO 2015 Written Page: 208 of 218 25 June 2015 ID: S96 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: BANK: ILO 2010 NRC EXAM Due to an error in the calibration procedure for t he RPS Drywell Pressure instruments, the high pressure trip setpoint for all four channels were adjusted such that the channels would not trip until Drywell Pressure reaches 2.2 psig.

In order to satisfy the required Technical Specif ications, readjust the trip setpoint for channels

__________________________.A.A and D to below Technical Specification limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

B.A and C to below Technical Specification limits within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

C.B and D to below Technical Specification limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

D.B and C to below Technical Specification limits within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .Answer:A ILO 2015 Written Page: 209 of 218 25 June 2015 Answer Explanation:

To restore trip capability for Drywell Pressure, at least one of the trip syst ems must be restored (trip setpoint less than 1.88 psig) within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . To restore a trip system the two channel s for that trip system

[(A or C) and (B or D)] must be restored.

Distracter Explanation:

Distracters are plausible and incorrect based on understanding trip setpoint and logic listed above.

Reference Information:

23.601 Pg 20 (Logic for Drywell Pressure - High)

T.S. 3.3.1.1 (setpoint and requirement)

NUREG 1123 KA Catalog Rev. 2G2.2.23Ability to track Technical Specif ication limiting conditions for operations.295024High Drywell Pressure.

Technical Specifications3.3.1.1Reactor Protection Syst em (RPS) Instrumentation 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

NRC Exam Usage 2015 ILO Exam Early Review Questions for NRC 2015 ILO ILO 2015 Written Page: 210 of 218 25 June 2015 ID: S97 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: MODIFIED BANK ID: 32686 The RHRSW System is designed to provide cooli ng water for the Residual Heat Removal (RHR)

System heat exchangers, required for a safe r eactor shutdown following a Design Basis Accident (DBA). Which ONE of the followi ng will ensure this design capacity?A.Div I RHR Reservoir at 26 feet, Div II Reservoir at 25.5 feet RHR Reservoir average temperature of 64°FB.Div I RHR Reservoir at 25.5 feet, Div II Reservoir at 25.2 feet RHR Reservoir average temperature of 81°FC.Div I RHR Reservoir at 25.4 feet, Div II Reservoir at 25.6 feet RHR Reservoir average temperature of 82°FD.Div I RHR Reservoir at 24 feet, Div II Reservoir at 25 feet RHR Reservoir average temperature of 77°FAnswer:A ILO 2015 Written Page: 211 of 218 25 June 2015 Answer Explanation:

RHR Service Water requires an adequate suction sour ce (Ultimate Heat Sink) bounded by a minimum independent and average level of 25 ft and temperature

<80°F with both towers and fans operable.

Distracter ExplainationB.Is incorrect and plausible because RHR Reservoir average temperature is too high. The examinee incorrectly focus on the Reservoir level only.C.Is incorrect and plausible because RHR Reservoir average temperature is too high. The examinee incorrectly focus on the Reservoir level only.D.Is incorrect and plausible because Average of both levels is less than the required 25 ft. The examinee incorrectly focus on the RHR Reservoir average temperature only.

NUREG 1123 KA Catalog Rev. 2G2.2.25Knowledge of bases in Technical Specific ations for limiting conditions for operations and safety limits.

Technical Specifications3.7.1 Residual Heat Removal Service Water (RHRSW) System3.7.2 Emergency Equipment Cooling Water (EECW)/Emergency Equipment Service Water (EESW) System and Ultimate Heat Sink (UHS) 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (2) Facility operating limitations in the technical specifications and their bases.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 212 of 218 25 June 2015 ID: S98 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: BANK: EQOP8320001A002001 Emergency Operations Facilities are NOT manned.Following an accident, it is required to estimate Core / Fuel Damage using the following Containment High Range Radiation M onitor (CHRRM) readings and conditions:*Reactor was SHUTDOWN at 1200.*CHRRM Readings were taken at 1300.

DIV 1 CHRRM indicates 2.0 x 10 4 R/hr.*DIV 2 CHRRM indicates 1.5 x 10 4 R/hr.Given EP-547 Attachment 1 and Enclosure B, which ONE of the following is the correct Core /

Fuel Damage calculation, based on these readings?

% Gap Release (H)% of Fermi-2

Upper Bound

LOCA (J) % of Regulatory

Guide 1.3 LOCA (K) A.B.

C.

D.Answer:B ILO 2015 Written Page: 213 of 218 25 June 2015 40.0 11.8 21.45.0 1.9 115.4 30.08.8 28.66.7 2.5 11.8 Answer Explanation:

Answer calculated using Enl. B 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months values for (E) 7E+4, (F) 3E+5, and (G) 8E+5 Plant Procedures EP-547 NUREG 1123 KA Catalog Rev. 2G2.3.5Ability to use radiation monitoring syst ems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (4) Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 214 of 218 25 June 2015 ID: S99 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW The plant was operating at 100% power. The Nort h RFP tripped. The following actions are in progress:*SBFW has been started and is injecting at 1200 gpm.*The P603 operator is inserting the CRAM array to lower Reactor Power.

Reactor water level is 185 inches and rising slowly.

Which ONE of the following describes (1) Reac tor Power, and (2) the proper followup actions directed by CRS?A.(1) 95%(2) Monitor core thermal limits.B.(1) 88%(2) Monitor core thermal limits.C.(1) 85%(2) Shutdown SBFW.D.(1) 65%(2) Shutdown SBFW.Answer:D ILO 2015 Written Page: 215 of 218 25 June 2015 Answer Explanation:

20.107.01 LOSS OF FEEDWATER OR FEEDWATER CO NTROL directs actions for a tripped RFP and RWL is low or lowering because a lack of Adequate pumping capacity.

CONDITION D. Reactor Power > 65% AND RFP tripped. (pg 5)

ACTION:

D.1 Verify RR runs back to 2/3 Limiter.

D.2 Place Recirc A & B Flow Limiter 2/3 Defeat switch in NORMAL.

D.3 Inject with SBFW at 1200 gpm.

D.4 Insert CRAM array to lower Reactor Power <

65%.Additionally when CONDITION E. CRAM array inserted. (pg 5)

ACTION:

E.1 Verify Reactor Power <

65%.E.2 Shutdown SBFW.

E.3 Monitor Core Thermal Limits.

E.4 Notify SNE.A.Is plausible and incorrect because they are different and incorrect power level / actions required by the Loss of Feedwater AOP. B.Is plausible and incorrect because they are different and incorrect power level / actions required by the Loss of Feedwater AOP.C.Is plausible and incorrect because they are differ ent and incorrect power level / actions required by the Loss of Feedwater AOP.

Ref (20.107.01 pg 3,5)

Plant Procedures 20.107.01 NUREG 1123 KA Catalog Rev. 2G2.4.11Knowledge of abnormal condition procedures

.295031Reactor Low Water Level.

10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 216 of 218 25 June 2015 ID: S100 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW The plant is operating at 100% power when, due to a loss of feedwater a reactor SCRAM occurs.

The Shift enters the EOPs on RPV water level 3, and takes appropriate actions.

20 minutes later, the plant is stable with R PV water level being maintained 173-214 inches, when 16D27 FIRE ALARM occurs. Fire Detection Z one 14/ Fire Zone 11ABW: Div 2 Battery Charger Room is alarming. The field operator reports that the fire door to Div 2 Battery Charger Room is hot to the touch and there is a strong odor of smoke in the area.

The CRS Enters 20.000.22 Plant Fires and begins di recting the shift to respond to the fire.

If the fire were to spread through the Div 2 Ba ttery Charger Room fire doors to the adjacent areas, which of the following actions would the CRS direct and why?A.PERFORM 20.000.18, Control Of The Plant From The Dedicated Shutdown Panel, because of a fire in a 3L zone.B.DIRECT the CRNSO to use HPCI and SBFW as preferred makeup sources, because the fire has or could impact RCIC.C.DIRECT the CRNSO to use RCIC and SBF W as preferred makeup sources, because the fire has or could impact HPCI.D.PERFORM 20.000.19 Shutdown From Outs ide The Control Room, because of a fire has damaged the ability of MCR to control safe shutdown equipment.Answer:A ILO 2015 Written Page: 217 of 218 25 June 2015 Answer Explanation:

The DC MCC Area is one of the immediately adjacent areas. Therefore:

Per 20.000.22 CONDITION C Fire in 3L Zone (8, 9A, 11, 12, 12A, 14 (DC MCC Area) or 16 (AB 4 only).

ACTION: C.1 PERFORM 20.000.18, Control Of T he Plant From The Dedicated Shutdown Panel B & C. Both HPCI and RCIC are compromised by the fire with spread to adjacent areas, Both of these are plausible if the examinee fails to underst and the relationships between HPCI, RCIC and ESF Divisional DC power. Additionally t hese distracters are quoted from the fire strategies listed in 20.000.22 Plant fires.

D. This answer is plausible if the examine does not understand the purpose of the AOP or confuses the purpose of this AOP with 20.000.18 Control Of The Plant From The Dedicated Shutdown Panel. Both AOPs do control the plant from outside the MCR.

Plant Procedures 20.000.22 NUREG 1123 KA Catalog Rev. 2 G2.4.27 3.4/3.9 Knowledge of "fire in the plant" procedure 10CFR55 RO/SRO Written Exam Content10 CFR 55.43(b) (5) Assessment of facility condi tions and selection of appropriate procedures during normal, abnormal, and emergency situations.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 218 of 218 25 June 2015 ID: R01 Points: 1.00Difficulty:

0.00 Level of Knowledge:

High Source: NEW The plant is operating with the following indications:

B21-R612 Jet Pump Total Flow Recorder 87.4 Mlbs/hr C51-R608A Rx Power APRM 1 100%B31-R621A/B RR MG SET SPEED CONTROLLERs 73% (controlling in AUTO)

Based on the above starting conditions and assu ming no operator action in response to the transient, opening the field breaker for 'A' Reac tor Recirc Pump will cause Reactor Power to lower. The power reduction is due to _______.A.'A' Reactor Recirc pump coasting down ONLY.

B.'A' Reactor Recirc Pump coasting down PLUS a speed reduction in 'B' Reactor Recirc pump because of #4 limiter.C.'A' Reactor Recirc Pump coasting dow n PLUS a speed reduction in 'B' Reactor Recirc pump because of #2 and #3 limiters.D.'A' Reactor Recirc Pump coasting dow n PLUS a speed reduction in 'B' Reactor Recirc pump because of limiter enforcem ent, placing the core in the SCRAM region.Answer:A ILO 2015 Written Page: 1 of 218 25 June 2015 Answer Explanation:

Limiter 4 is the only runback that would actuate.

At a setting of 75%, it would have no effect on the reduction on flow. Therefore the coast down of A pump is the only thing lowering Rx power.

Distractor Explanation:B.This answer is incorrect because the setpoint of Limiter #4 is 75%, this answer is plausible if the examinee incorrectly thought #4 lim iter as a lower setpoint.C.This answer is incorrect and because Limiters #2 and

3 are cut out. This answer is plausible if the examinee incorrectly believes #2 & #3 limiters will enforce. D.This answer is incorrect because Limiter #2 and #3 are cut out, this answer is plausible if the examinee incorrectly believes flow would lower due to limiter enforcement.

Reference Information:

23.138.01 Section 1.0 Description (Limiter

2,#3,#4 on page 8) - explains limiters Plant Procedures 23.138.01 NUREG 1123 KA Catalog Rev. 2295001 AK3.Knowledge of the reasons for the followi ng responses as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION :295001 AK3.02Reactor power response 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam Early Review Questions for NRC 2015 ILO ILO 2015 Written Page: 2 of 218 25 June 2015 ID: R02 Points: 1.00Difficulty:

0.00 Level of Knowledge:

Low Source: NEW Following a loss of 120 kV and a reactor SCRAM, the following indications exist on panel H11-P809: Breaker Positions BUS 64B POS B8 CLOSED BUS 64B POS B6 TRIPPED BUS 11EA POS EA3 CLOSED BUS 12EB POS EB3 OPEN BUS 64C POS C8 CLOSED BUS 64C POS C6 TRIPPED Based on these indications (1) what event has occurred, and (2) what actions should the operating crew take in addition to entering AOP 20.300.120kV?A.(1) Bus lockout on BUS 64B (2) Perform 20.300.72B, Loss Of Bus 72B ONLY.B.(1) Bus lockout on BUS 64C (2) Perform 20.300.72C, Loss Of Bus 72C ONLY.C.(1) EDG 12 failure due to a fault with EDG 12 (2) Perform 20.307.01, Emergency Diesel Generator Failure AND 20.300.72C, Loss Of Bus 72C.D.(1) EDG 11 failure due to a fault with EDG 11 (2) Perform 20.307.01, Emergency Diesel Generator Failure AND 20.300.72B, Loss Of Bus 72B.Answer:C ILO 2015 Written Page: 3 of 218 25 June 2015 Answer Explanation:

20.300.120kV directs 20.307.01 and 20.300.72C if the EDG 12 output breaker fails to close (Condition K). Additionally, no bus lockout is indicated so 20.

307.01 would be effective if the cause of the EDG failure is correctable.

Distracter Explanation:A.Is incorrect and plausible because there is no i ndication of a 64B BUS LOCKOUT (indicated by BUS isolation). If there was a 64B BUS LOCK OUT, entering 20.300.72B would be correct.B.Is incorrect and plausible because there there is no indication of a 64C BUS LOCKOUT (indicated by BUS isolation). If there was a 64C BUS LO CKOUT, entering 20.300.72C would be correct.D.Is incorrect and plausible because the BUS 11EA POS EA3 is CLOSED so EDG 11 has started and loaded. If EDG 11 had failed then 20.307.

01 and 20.300.72B would be correct.

Reference Information:

20.300.120kv, page 9, Condition K.

Plant Procedures 20.300.120kv NUREG 1123 KA Catalog Rev. 2295003 AA1. Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER :

295003 AA1.02 Emergency generators 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 4 of 218 25 June 2015 ID: R03 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW-CONTRACT At 1030, a loss of offsite power occurred. AOP 20.300.Offsite, Loss of Offsite Power, is being implemented to mitigate the electrical system event.

At 1145 (current time), electrical system indications and conditions are as follows: *All EDGs are running loaded. *120kV and 345kV switchyard components are damaged and are being evaluated by personnel for restoration. *10D72 BOP 260/130V BATTERY 2PC TROUBLE is in alarm.

9D17 DIV I ESS 130V BATTERY 2PA TROUBLE alarmed but is now clear. *10D68 DIV II ESS 130V BATTERY 2PB TROUBLE alarmed but is now clear.

If the electrical system is operating within its design bases, based on the indications, (1) what is the status of the station batteri es and (2) what is their ability to supply loads during the event?A.(1)Battery 2PC is below minimu m cell voltage, loads are de-energized.(2)Batteries 2PA and 2PB are above minimum cell voltage, loads are supplied adequate power to operate.B.(1)Battery 2PC is above minimum cell voltage, BOP loads are supplied adequate power to operate.(2)Batteries 2PA and 2PB are above minimum cell voltage, ESF loads are supplied adequate power to operate.C.(1)Battery 2PC is above minimum cell voltage, BOP loads should be stripped in order to supply adequate power to remaining loads.(2)Batteries 2PA and 2PB are above minimum cell voltage, ESF loads are supplied adequate power to operate.D.(1)Battery 2PC is below minimum cell voltage, BOP loads should be stripped in order to supply adequate power to remaining loads.(2)Batteries 2PA and 2PB are below mi nimum cell voltage, ESF loads should be stripped in order to supply adequate power to remaining loads.Answer:C ILO 2015 Written Page: 5 of 218 25 June 2015 Answer Explanation:

Based on the design of the BOP batteries, under loaded conditions the batteries can supply loads for up to 90 minutes before falling below minimum cell vo ltage with no charger available. Based on the conditions of the 120kv switchyard, currently no BO P charger is available. The procedure requires stripping the DC BOP busses after one hour if the DC BOP chargers remain unavailable. This is to retain the batteries available for DC control power use up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . ESF batteries are fully available with chargers energized due to the EDG power being availabl

e. Alarms indicate battery chargers are in service for the ESF batteries as directed by the procedure.

Distracter Explanation:A.Is incorrect and plausible because the examinee c ould incorrectly interpret the BOP battery trouble alarm as a loss of the battery voltage. The volt age alarm is above the minimum cell voltage. This would also be supported by the indication that o ffsite power cannot be restored. ESF chargers are energized as indicated by the alarm status whic h also enhances the plausibility of the distracter. B.Is incorrect and plausible because the examinee could incorrectly interpret that since EDG power is available all available battery chargers bot h ESF and BOP are energized. ESF chargers are energized as indicated by the alarm status whic h also enhances the plausibility of the distracter. D.Is incorrect and plausible because the examinee could incorrectly interpret the indication and application of the knowledge of the design basis of

1.5 hour5.787037e-5 days 0.00139 hours 8.267196e-6 weeks 1.9025e-6 months s

to the BOP battery and conclude that since one hour has passed without stripping the BO P battery, that the cells are below minimum voltage.Reference Information:

R32-00 DC ELECTRICAL SYSTEM DBD pages 26-27 Criteria for BOP batteries Plant Procedures 20.300.Offsite R32-00 DESIGN BASIS DOCUMEN T FOR DC ELECTRICAL SYSTEM NUREG 1123 KA Catalog Rev. 2295004 AA2.Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER :295004 AA2.03Battery voltage 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam

Early Review Questions for NRC 2015 ILO ILO 2015 Written Page: 6 of 218 25 June 2015 ID: R04 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: BANK:802-2003-0006-008 With the plant operating at 90% power, a F eedwater Controller Failure - Maximum Demand occurs. Automatic protection from this transient is assured by a scram generated as a result of

________.A.OPRM - Upscale direct scram, whic h prevents exceeding the MCPR Safety LimitB.Reactor Water Level High - Level 8 Main Turbine Trip indirect scram, which prevents exceeding the MCPR Safety LimitC.Reactor Vessel Steam Dome Pressure - High direct scram, which prevents exceeding the Reactor Coolant System Pressure Safety LimitD.APRM Simulated Thermal Power Upscale direct scram, which prevents exceeding the Reactor Coolant System Pressure Safety LimitAnswer:B ILO 2015 Written Page: 7 of 218 25 June 2015 Answer Explanation:

The feedwater and main turbine high water level tr ip instrumentation is assumed to be capable of providing a turbine trip in the design basis transient analysis for a feedwater controller failure, maximum demand event. The Level 8 trip indirectly initiate s a reactor scram (above 30% RTP) from the main turbine trip and trips the feedwater pumps, thereby te rminating the event. The reactor scram mitigates the reduction in MCPR.

Distracter Explanation:A.This is plausible and incorrect because this prot ects against thermal hydr aulic power oscillations.C.This is plausible and incorrect because this prot ects against any event that result in a reactor pressure increase.D.This is plausible and incorrect because this prov ides protection against a loss of feedwater heating and ensures the MCPR SL is not exceeded.

Reference Information:T.S.B. 3.3.2.2 (pg 3.3.2.2-2) APPLICABLE SAFETY ANALYSES NUREG 1123 KA Catalog Rev. 2G2.2.38Knowledge of conditions and lim itations in the facility license.295005 AK1Knowledge of the operational implications of the following concepts as they apply to MAIN TURBINE GENERATOR TRIP:295005 AK1.02 Core thermal limit considerations.

Technical Specifications3.3.2.2 Feedwater and Main Turbine High Water Level Trip Instrumentation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 8 of 218 25 June 2015 ID: R05 Points: 1.00Difficulty:

3.50 Level of Knowledge:

High Source: BANK: 315-0048-A018-001 The plant was operating at 15% power during a startup. Due to an electrical malfunction, ALL Turbine Control Valves opened resulting in a Reactor Scram. No operator actions have been taken.Which one of the following lists the systems av ailable for Decay Heat Removal immediately following the scram?A.Safety Relief Valves and Reactor Water Cleanup ONLY.

B.Turbine Bypass Valves, Safety Relief Valves and Reactor Water Cleanup ONLY.C.Main Steam Line Drain Valves, Turbine Bypass Valves and Safety Relief Valves ONLY.D.Main Steam Line Drain Valves, Turbine Bypass Valves, Safety Relief Valves and Reactor Water Cleanup.Answer:A ILO 2015 Written Page: 9 of 218 25 June 2015 Answer Explanation:

With power at 15%, the Mode Switch is in RUN.

TCVs opening lowers RPV Pressure. When pressure reaches 756 psig with Mode Switch in RUN, a GP1 PCIS isolation occurs and a Scram. With MSIVs closed, only SRVs and RWCU are immediately available to remove decay heat.

Distracter Explanation:B. Is incorrect and plausible because Tu rbine Bypass Valves are not available.C.Is incorrect and plausible because Turbine By pass Valves and MSL Drains are not available.D.Is incorrect and plausible because Turbine By pass Valves and MSL Drains are not available.

Reference Information:

23.601 Main Steam Line Pressure - Low 23.601 Main Steam Line Isolation Valve - Closure Limit Switches

20.000.21 Bases Mode Switch interlocks Plant Procedures 20.000.21 20.000.21 Bases NUREG 1123 KA Catalog Rev. 2295006 AK1.Knowledge of the operational implications of the following concepts as they apply to SCRAM :295006 AK1.01Decay heat generation and removal 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 10 of 218 25 June 2015 ID: R06 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW-CONTRACTThe plant was operating at 100% power when a fire occurred in the Owner Controlled Area. The fire engulfed several buildings in cluding the onsite vehicle garage.

The following timeline of events subsequently occurred:10:30Heavy black smoke is drawn into the ventilation intake making the Control Room uninhabitable.11:15All offsite power is lost.

Based on the above plant conditions, at 10:30 cont rol of the plant would have shifted to the

__________________.A.Remote Shutdown Panel and remained there for the duration of the eventB.Dedicated Shutdown Panel and remained there for the duration of the eventC.Remote Shutdown Panel and subsequently to the Dedicated Shutdown Panel at 11:15D.Dedicated Shutdown Panel and subsequent ly to the Remote Shutdown Panel at 11:15Answer:C ILO 2015 Written Page: 11 of 218 25 June 2015 Answer Explanation:

Initially at 10:30 the correct location for remote cont rol of the plant due to an uncomplicated control room evacuation is the Remote Shutdown Panel. At time 11:15 control of the plant is required to shift to the Dedicated Shutdown Panel since the Remote Shutdown System is designed for an uncomplicated control room evacuation event and as such does not serve to mitigate the full range of design basis events.

Therefore, transfer to the Dedicated Shutdown Panel is required since this system provides a means of restoring AC power to the Div 1 ESF and BOP busse s as well as the necessary system controls and instrumentation to maintain the reactor core co vered in the event of a Control Room evacuation coincident with a loss of offsite power.

Distracter Explanation:A.Is incorrect and plausible because the examinee coul d incorrectly assume that the Remote Shutdown Panel was designed for this type of event. The candi date could fail to recognize that the Remote Shutdown System is designed for a control room ev acuation event (short term duration) and as such does not serve to mitigate the full range of design bas is events for which the Main Control Room was designed. Furthermore the candidate could also fa il to recognize that the Dedicated Shutdown System provides a means of restoring AC power to the Div 1 ESF and BOP busses. The panel also provides necessary system controls and instrumentation to maintain the reactor core covered. Power restoration may become necessary following a fire or loss of offsite power.B.Is incorrect and plausible if the examinee incorrect ly assumed that control of the plant from the Dedicated Shutdown Panel was required for all Main Control Room evacuation events that are a result of an onsite fire. The candidate could fail to recognize that control fr om the Dedicated Shutdown Panel is only required for fires in a 3L "zone of conc ern" that has the potential to impact safe plant shutdown capability due to multiple hot shorts and/or open circuits causing loss of control from the control room. D.Is incorrect and plausible if the examinee incorrect ly assumed that control of the plant from the Dedicated Shutdown Panel was required for all Main Control Room evacuation events that are a result of an onsite fire. The candidate could t hen conclude that, upon Loss of Offsite Power, the correct response would be to shift control to the Remote Shutdown Panel. The candidate could fail to recognize that the Dedicated Shut down System is designed to allow for remote control of the plant concurrent with a loss of offsite power event.

Reference Information:

20.000.19 (pg 2) Overide 1 20.000.19 BASES (pg 2) Overide 1 Bases

20.000.18 BASES (pg 2) - BASES

SUMMARY

ILO 2015 Written Page: 12 of 218 25 June 2015 Plant Procedures 20.000.18 20.000.19 20.000.18 Bases

20.000.19 Bases NUREG 1123 KA Catalog Rev. 2295016 AK2.Knowledge of the interrelations between CONTROL ROOM ABANDONMENT and the following:295016 AK2.02Local control stations: Plant-Specific 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 13 of 218 25 June 2015 ID: R07 Points: 1.00Difficulty:

2.00 Level of Knowledge:

High Source: NEW-CONTRACT The plant is operating at 100% power with the following conditions: *#2 & #5 GSW Pumps are running. *#3 GSW Pump is OOS for motor replacement.

1. 4 GSW Pump has just tripped.

1. 6 GSW Pump is OOS for discharge strainer leak repair.

GSW Header Pressure steady at 70 psig with the P4100-F841, GSW Bypass Line Pressure Ctrl Vlv, 50% open. *RBCCW Heat Exchanger outlet temperature is 90°F and rising.

Which ONE of the following actions is requi red to stabilize the above conditions?A.Dispatch an operator to throttl e closed P4100-F841, GSW Bypass Line Pressure Ctrl Vlv, in order to raise GSW header pressure and restore adequate cooling water flow.B.Dispatch an operator to throttle open P4100-F840, GSW Flow Test Pressure Ctrl Vlv, in order to raise GSW header pressure and restore adequate cooling

water flow.C.Scram the reactor, trip the main turb ine and initiate Div 1 and 2 EECW in order to establish cooling to safety related equipment and restore GSW header

pressure to normal band.D.Increase cooling water flow using P 42-F400, RBCCW Temp Control Vlv in AUTO, or MANUAL if necessary, in or der to restore RBCCW Heat Exchanger outlet temperature and GSW header pressure in band.Answer:A ILO 2015 Written Page: 14 of 218 25 June 2015 Answer Explanation:

P4100-F841, GSW Bypass Line Pressure Ctrl Vlv is a dum p, or backpressure, control valve. Prior to the transient and trip of #4 GSW Pump, this valve was required to be open to maintain GSW header pressure due to 3 pumps being in excess of the plant cooling water requirements. With only 2 GSW pumps running, closing this valve will raise GSW header pre ssure and increase cooling water flow to the CCW heat exchangers cooled by GSW thereby lowering CCW cooling temperatures, such as the elevated

RBCCW Heat Exchanger outlet.

Distracter Explanation:B.Is incorrect and plausible because the examinee could incorrectly determine that, since the P4100-F841, GSW Bypass Line Pressure Ctrl Vlv, is at its 50% open limit then pressure control would logically transfer to the P4100-F840, GSW Flow Test Pressure Ctrl Vlv. This would be correct for a high pressure condition. The examinee could also incorrectly determine that GSW pressure control valves are throttled OPEN to raise GSW header pressure, however, the valves are actually backpressure control valves and must be throttled CLOSED to raise GSW header pressure and increase cooling water flow to supported CCW heat exchangers.C.Is incorrect and plausible because the candidate could incorrectly determine that, since GSW header pressure is below 65#, the plant and main turbine should be tripped, and EECW initiated, as directed

by the loss of GSW AOP. However, the loss of GSW AOP override statement only requires a plant scram if GSW header pressure cannot be restored AND MAINTAINED above 65 psig. The examinee should determine that throttling capacity exists to restore GSW header pressure without the need for these drastic actions.D.Is incorrect and plausible because the candidate c ould incorrectly determine that throttling open the P42-F400, RBCCW Temp Control Vlv in AUTO, or MANUAL, would corre ct the elevated RBCCW Heat Exchanger outlet temperature condition.

Although this action is directed by ARP 2D120, RBCCW HX DISCH TEMPERATURE HIGH/LOW for a High Temperature condition, the action is accompanied by a conditional statement to ma ke the adjustments while monitoring GSW header pressure. With GSW header pressure alr eady degraded, this action would not be prudent.

Reference Information:

ARP 7D14 (pg 1) GSW pressure low actions

23.131 (pg 23-25) GSW pressure control ILO 2015 Written Page: 15 of 218 25 June 2015 Plant Procedures 23.131 07D14 23.127 02D120 NUREG 1123 KA Catalog Rev. 2295018 AK3.Knowledge of the reasons for the followi ng responses as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER:295018 AK3.06Increasing cooling water flow to heat exchangers.

10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (5) Facility operating characterist ics during steady state and transient conditions, including coolant chemistry, causes and effects of temperature, pressure and reactivity changes, effects of l oad changes, and operating limitations and reasons for these operating characteristics.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 16 of 218 25 June 2015 ID: R08 Points: 1.00Difficulty:

3.00 Level of Knowledge:

High Source: NEW The plant is operating at 100% power with the following Station Air Compressor lineup:East Station Air Compressor Off Center Station Air Compressor Running West Station Air Compressor Auto Following a seismic event, Bus 72A has been de-energized, and both Control Air Compressors auto start due to lowering air pressure. At 2 ps i below the Control Air Compressors auto start setpoint, all air header pressures begin to recove

r. Assuming no operator action, what is the status of the following Station Air valves?P50-F401P50-F402P50-F440 STATION AIR

TO TB HDR ISO VLV STATION AIR

TO NIAS ISO VLV DIV 1 CONTROL

AIR ISO VLV A.B.

C.

D.Answer:B ILO 2015 Written Page: 17 of 218 25 June 2015 CLOSED CLOSED OPEN OPEN OPEN CLOSED CLOSED OPEN CLOSED OPEN OPEN CLOSED Answer Explanation:

Based on logic prints listed & AOP 20.129.01, Control Air Compressors start at 85 psi therefore system press has to go to below 85 psi and then, per the questi ons stem pressure recovers. Based on this the 401 will get a closed signal, and the 402 and 440 will not get close signals.

Distracter Explanation:

Distracter are plausible based on not understanding setpoints or power supplies/auto-starts of compressors A.Is incorrect because 401 will get close signal at 85#.

This answer is plausible if the examinee incorrectly assumes the system pressure did not go low enough to cause isolations.C.Is incorrect because 402 will not get close signal until 75# and air header only went to 83#. This answer is plausible if the examinee incorrectly assumes the system pressure did go low enough to cause isolations.D.Is incorrect because 402 & 440 will not get close signal until 75# and air header only went to 83#.

This answer is plausible if the examinee incorre ctly assumes the system pressure did go low enough to cause isolations.

Reference Information:

I-2450-02,04,05 are not shown here, however they ar e included in development folder, the AOP actions match the actuations.

20.129.01 (pg 3) System actuation bas ed on pressure per action statement.

Plant Procedures 20.129.01 20.300.72A NUREG 1123 KA Catalog Rev. 2295019 AA1.Ability to operate and/or monitor the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR:295019 AA1.04Service air isolations valves: Plant-Specific 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

NRC Exam Usage 2015 ILO Exam ILO 2015 Written Page: 18 of 218 25 June 2015 ID: R09 Points: 1.00Difficulty:

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High Source: MODIFIED BANK: ID 26798 The plant is in HOT SHUTDOWN with the following conditions:*Reactor Pressure is 25 psig.*RPV Water Level is 250 inches and stable.

RHR Pump B is operating in Shutdown Cooling Mode.

When the following alarm occurs:*2D26, DIV II RHR SYSTEM LOW FLOW BYPASS INITIATED.

Without operator action which one of the followi ng describes (1) the effect on Reactor Water Level and (2) reason for that effect?A.(1) Reactor Water Level will remain the same.

(2) The RHR pumps min flow valve has opened; system flow rate is reduced.B.(1) Reactor Water Level will remain the same. (2) The RHR system flow is low, indicating a low heat load condition.C.(1) Reactor Water Level will rise. (2) The RHR system flow from the vessel is low, this indicates a flow imbalance.D.(1) Reactor Water Level will lower.

(2) The RHR pumps min flow valve has opened, rejecting Rx water to the Torus.Answer:D ILO 2015 Written Page: 19 of 218 25 June 2015 Answer Explanation:

The alarm 2D26 indicates the RHR pumps min flow valve has opened resulting in increased pump flow (two paths). The min flow valve will reject Rx wate r to the Torus causing level to lower until a low Rx water level isolation is reached.

OE: SOER 87-002, Inadvertent Draining of the Reactor Vessel to Suppression Pool at BWRs Distracter Explanation:A.This is plausible and incorrect because SDC is a closed loop system, however when the minflow valve is open it becomes a rejection path causing level to lower.B.This is plausible and incorrect because the examinee could identify the low flow alarm to be low flow in a closed loop system, which would have no effect on level.C.This is plausible and incorrect because the exam inee could identify the low flow alarm with RHR output from the vessel, and given a feed and bleed type system this could identify a condition that

causes level to rise.

Reference Information:

ARP 2D26 (pg 1) Action for this condition.

FOS M-5706-1 Green highlight of path for drain to torus (located in Development folder)

Plant Procedures 03D079 02D026IERs/ SOERs/ SERs/ SENs/ O&MR SOER 87-2 Inadvertant Draining of Reac tor Vessel to Suppression Pool at BWRs NUREG 1123 KA Catalog Rev. 2295021 AA2.Ability to determine and/or interpret the following as they apply to LOSS OF SHUTDOWN COOLING:295021 AA2.03Reactor water level 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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High Source: NEW-CONTRACT Core alterations are in progress with the plant in Mode 5. 'A' RHR pump is operating in the shutdown cooling mode and 'C' RHR pump is ava ilable in standby. 'B' and 'D' RHR pumps are out of service. The control room receives a report that spent fuel pool level is slowly lowering. A nuclear operator isolates the source of the leak and the level reduction is stopped. *2D1 Fuel Pool Level Low is alarmed.*2D13 Fuel Pool Cooling Trouble is alarmed.

FPCCU Pumps have tripped.

The control room determines that RPV level is 20.0 ft above the RPV flange using B21-R605, RPV Water Level Flood Up Range.

Which one of the following actions is required?A.Suspend movement of fuel assemblies within the RPV.B.Verify an alternate method of decay heat removal is available.

C.Verify two alternate methods of decay heat removal are available.

D.Initiate action to restore sec ondary containment to operable status.Answer:A ILO 2015 Written Page: 21 of 218 25 June 2015 Answer Explanation:

The examinee should correctly determi ne that Technical Specification 3.9.6 (Refueling Operations - RPV Water Level) requires suspension of movement of f uel assemblies with RPV level < 20 ft 6 inches. The completion time for the action is IMMEDIATELY.

Distracter Explanation:

B.Is incorrect and plausible because examinee may in correctly identify Techni cal Specification 3.9.7 actions require one verification that an alternate method of decay heat removal is available with a completion time of IMMEDIATELY if one RHR SDC subs ystem is not in operation. With level < 20 ft 6 inches, this LCO is not applicable.C.Is incorrect and plausible because the examinee c ould incorrectly identify Technical Specification

3.9.8 requires

verification of an alternate met hod of decay heat is available for each inoperable required RHR SDC subsystem. Although B and D are inoperable two still remain and even though RPV level is less than 20 ft 6 inches which makes the LCO applicable, the condition for entry is not currently met.D.Is incorrect and plausible because the examinee c ould incorrectly identify Technical Specification 3.9.7 and 3.9.8 have actions to IMMEDIATELY rest ore secondary containment to operable status if

the completion times of condition A in both LCOs are not met.

Reference Information:

T.S. 3.9.6 (pg 3.9-9) LCO and Immediate action.

NUREG 1123 KA Catalog Rev. 2G2.2.39Knowledge of less than one hour techni cal specification action statements for systems.295023Refueling Accidents Technical Specifications3.9.6 Reactor Pressure Vessel (RPV) Water Level 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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Low Source: NEW-CONTRACT Which of the responses below best completes t he following regarding the operational implications of High Drywell Pressure?

In MODES 1, 2 and 3, the maximum allowable Dryw ell Pressure is __(1)__. This limit is based on __(2)__.A.(1) 2.25 psig (2) maintaining the resultant differential pressure below the maximum primary containment design differential pressure in the event of inadvertent drywell spray actuation.B.(1) 2.0 psig (2) maintaining the resultant differential pressure below the maximum primary

containment design differential pressure in the event of inadvertent drywell spray actuation.C.(1) 2.25 psig(2) maintaining the resultant peak primary containment accident pressure below

the primary containment design pressu re in the event of a Design Basis Accident (DBA).D.(1) 2.0 psig(2) maintaining the resultant peak primary containment accident pressure below

the primary containment design pressu re in the event of a Design Basis Accident (DBA).Answer:D ILO 2015 Written Page: 23 of 218 25 June 2015 Answer Explanation:

The examinee should correctly conclude that, in a ccordance with LCO 3.6.1.4, the maximum Drywell Pressure in the specified MODEs is 2.0 psig. The examinee should also conclude that the bases for this limit is to preserve the initial conditions assumed in the accident anal ysis for the Design Basis Accident (DBA) thus ensuring that the peak primary containment pressure does not exceed the design pressure of the primary containment pressure.

Distracter Explanation:A.Is incorrect and plausible because the examinee c ould fail to recognize the LCO 3.6.1.4 maximum allowable Drywell Pressure of 2.0 psig. Also, t he examinee could fail to recognize that the basis given in the second part of the distractor is for t he minimum allowable Drywell Pressure specified by LCO 3.6.1.4 in MODES 1, 2 and 3, which protects containment integrity by keeping external to internal drywell differential pressure below the maximum allowable design D/P.B.Is incorrect and plausible because the first part of the distractor does provide the correct maximum Drywell Pressure allowed by LCO 3.6.1.4 in the MO DEs specified. However, the examinee could fail to recognize that the basis given in the second par t of the distractor is for the minimum allowable Drywell Pressure specified by LCO 3.6.1.4 in MODES 1, 2 and 3, which protects containment integrity by keeping external to internal drywell differential pre ssure below the maximum allowable design D/P.C.Is incorrect and plausible because the examinee c ould fail to recognize the LCO 3.6.1.4 maximum allowable Drywell Pressure of 2.0 psig.

Reference Information:

T.S. 3.6.1.4 (pg 3.6-18) LCO statementT.S. 3.6.1.4 BASES (pg B3.3.1.4-1 to 2) Bases for limit.

Plant Procedures 03D081 NUREG 1123 KA Catalog Rev. 2295024 EK1.Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL PRESSURE:295024 EK1.01Drywell integrity: Plant-Specific Technical Specifications3.6.1.4Primary Containment Pressure 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.

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Low Source: NEW-CONTRACT The reactor is near the end of an operating cycl e with Power and Recirculation Flow at 100%.

If the plant responds as assumed in the Fermi 2 over-pressurization protection analysis, when at least two Average Power Range Monitors reach

________ , the reactor will scram in order to terminate a Main Steam Isolation Valve (MSIV) closure event.A.107.2% Neutron FluxB.113.5% Simulated Thermal Power C.118% Neutron Flux D.122.2% Simulated Thermal PowerAnswer:C ILO 2015 Written Page: 25 of 218 25 June 2015 Answer Explanation:

As the MSIV closure event is occurring, reactor pressure will be increasing. The increase in RPV pressure during reactor operation will compress the steam voids and result in a positive reactivity insertion, which then causes neutron flux and Thermal Power to increase. The Fermi 2 over-pressurization protection analysis conservatively assumes scram on the Average Power Range Monitor Neutron Flux-Upscale trip of 118% which, along with the SRVs, limits the peak RPV pressure to less than the ASME Code limits.

Distracter Explanation:A. Is incorrect and plausible if the candidate a ssumed that the 107.2% was the APRM setpoint. The 107.2% is the RBM HTSP at >82% reactor power which will only cause a Rod Withdrawl Block.B.Is incorrect and plausible because the examinee c ould fail to recognize that the APRM STP-Upscale trip provides protection against tr ansients where Thermal Power increases slowly and not for rapidly changing power events such as the MSIV closure and, therefore, assumes that since the APRM STP-Upscale trip is clamped at an upper limit of 113.

5%, which is lower than the APRM Neutron Flux-Upscale Function Allowable Value at 100% recirculat ion flow, that this function will terminate the MSIV closure event before the APRM Upscale Function.D. Is incorrect and plausible because the examinee c ould fail to recognize that the APRM STP-Upscale trip provides protection against tr ansients where Thermal Power increases slowly and not for rapidly changing power events such as the MSIV closure and then calculates the STP upscale trip for 100%

recirculation flow using the following formula: 0.62(W-W) + 60.2%.

Reference Information:

T.S.B. 3.3.1.1 (pg B 3.3.1.1-14 to 15) Bases for MSIC Closure Plant Procedures 03D093 03D073 NUREG 1123 KA Catalog Rev. 2295025 EK2.Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following:295025 EK2.09Reactor power Technical Specifications3.3.1.1Reactor Protection Syst em (RPS) Instrumentation 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (7) Design, components, and functi on of control and safety systems, including instrumentation, signals, interlo cks, failure modes, and automatic and manual features.

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Low Source: NEW-CONTRACT Which of the responses below best completes the following statement regarding the critical parameter for the Heat Capacity Limit (HCL) curve and the basis behind why the reactor must be Emergency Depressurized when the curve is exceeded?

The reactor is Emergency Depressurized when

__(1)__ exceeds the HCL to ensure that the

__(2)__ is not exceeded before the depressu rization of the RPV is complete.A.(1) Drywell Temperature (2) Drywell Design TemperatureB.(1) Drywell Temperature (2) Pressure Suppression PressureC.(1) Torus Water Temperature (2) Torus Design TemperatureD.(1) Torus Water Temperature (2) Pressure Suppression PressureAnswer:C ILO 2015 Written Page: 27 of 218 25 June 2015 Answer Explanation:The critical parameter on the HCL curve is Torus Water Temperature. Emergency Depressurizing due to exceeding the HCL curve is based on not exceeding either the torus design temperature or the PCPL

before the depressurization of the RPV is complete.

Distracter Explanation:A.Is incorrect and plausible because the examinee c ould determine that the cr itical parameter on the HCL curve is Drywell Temperature and therefore conclude that Emergency Depressurizing due to exceeding the HCL curve is based on not exceeding the drywell design temperature before the

depressurization of the RPV is complete.B. Is incorrect and plausible because the examinee c ould determine that the cr itical parameter on the HCL curve is Drywell Temperature. The candidate could incorrectly conclude that Emergency

Depressurizing due to exceeding the HCL curv e is based on not exceeding the Pressure Suppression Pressure of containment rather than the PCPL limit before the depressurization of the RPV is complete.D.Is incorrect and plausible because the examinee could correctly recognize that the critical parameter on the HCL curve is Torus Water Temperature. However, the candidate could incorrectly conclude that Emergency Depressurizing due to exceeding the HCL curve is based on not exceeding the

Pressure Suppression Pressure of containm ent rather than the PCPL limit before the depressurization of the RPV is complete.

Reference Information:

BWROG EPGs/SAGs, Appendix B (pg B-7-24 & B-17-16 to 17) Bases of HCL & Emergency Depressurization.

Plant Procedures 29.100.01 SH 2 BWROG EPG App B NUREG 1123 KA Catalog Rev. 2295026 EK3.Knowledge of the reasons for the following responses as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:295026 EK3.01Emergency/normal depressurization 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b) (8) Components, capacity, and functions of emergency systems.

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High Source: NEW-CONTRACT At which of the following is the Drywell Tem peratures and RPV pressures may PRV Water level indication become unreliable due to boiling in the instrument run? (See attached 29.100.01 Sheet 6.)Drywell TemperatureRPV pressureA.250°F25 PSIGB.325°F50 PSIG C.350°F150 PSIG D.375°F250 PSIGAnswer:B ILO 2015 Written Page: 29 of 218 25 June 2015 Answer Explanation:

Using the RPV Saturation Temperature curve, at 50 ps ig in the Reactor, the Drywell Temperature listed at which the instrument may become reliable due to boiling in the run is aboove 300°F.

Distracter Explanation:

Distracters a incorrect and plausible based in plotti ng the data in the RPV Saturation Temperature curve.

Reference Information:

29.100.01 SH 6 not include here, a copy is in the handout folder.

Plant Procedures 29.100.01 SH 6 NUREG 1123 KA Catalog Rev. 2295028 EA2.Ability to determine and/or interpret t he following as they apply to HIGH DRYWELL TEMPERATURE:295028 EA2.02Reactor pressure 10CFR55 RO/SRO Written Exam Content10 CFR 55.41(b)(10) Administrative, normal, abnormal, and emergency operating procedures for the facility.

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High Source: NEWA plant event has damaged the Reactor Building. The Operating Shift is executing the EOPs.*The reactor has scrammed however 12 rods are stuck at 04*RWL is 176 inches.

Reactor Pressure is 900 psig.

The MSIVs are open.The STA has plotted three points on the PSP based on valid data from MCR indication.