ML110200547

Draft - Written Operator Licensing Exam (Folder 2)

ML110200547
Person / Time
Site:	Susquehanna
Issue date:	10/26/2010
From:	Susquehanna
To:	Caruso J Operations Branch I
Hansell S
Shared Package
ML102180025	List: IR 05000387/2011301 ML102530213 ML102650461 ML110070539 ML110130101 ML110200472 ML110200500 ML110200547 ML110200558 ML110400227 ML110400252 ML110400267 ML110400275 ML110811006
References
TAC U01793
Download: ML110200547 (200)

Text

QUESTION 1 Both Units are operating at full power.

The electrical distribution system is in a normal full power line up, EXCEPT that breaker 2A201 09, Alternate Supply to ESS bus 2A, is INOPERABLE and is RACKED OUT for maintenance.

Breaker OA10306, Startup Bus 10 feeder to XFMR-1 01, then TRI PS UNEXPECTEDLY due to a breaker mechanism failure.

Which one of the following describes the response (if any) to this breaker failure?

A.

• ESS Bus 1A will REMAIN ENERGIZED from its NORMAL supply;

• ESS Bus 2A will be RE-ENERGIZED from Diesel Generator A when the DG exceeds 540 RPM and 90% rated voltage.

B.

• ESS Bus 1A will be RE-ENERGIZED from its ALTERNATE supply;

• ESS Bus 2A will REMAIN ENERGIZED from its NORMAL supply.

C.

• ESS Bus 1A will REMAIN ENERGIZED from its NORMAL supply;

• ESS Bus 2A will REMAIN ENERGIZED from its NORMAL supply.

D.

• ESS Bus 1A will be RE-ENERGIZED from its ALTERNATE supply;

• ESS Bus 2A will be RE-ENERGIZED from Diesel Generator A when the DG exceeds 540 RPM and 90% rated voltage.

LOC-23 NRC Exam Rev 3

K&A # 264000 K1.01 Importance Rating 3.8 QUESTION 1 RO Tier 2 Group 1 K&A Statement: Knowledge of the physical connections and/or cause- effect relationships between EMERGENCY GENERATORS (DIESEUJET) and the following: A.C. electrical distribution Justification:

A. Incorrect, Bus 1A will auto transfer to its alternate supply, S/U bus 20 when its normal supply, bus 10 is de-energized by the breaker failure. Candidates may select this if they understand that bus 2A will be energized by DG A after it reaches rated speed and voltage, but fail to correctly recall the normal and alternate supplies to bus 1A.

B. Incorrect, the normal supply to both busses 1A and 2A is lost when S/U bus 10 de energizes. Candidates may select this if they do not recognize that the normal power supply to bus 2A is lost.

C. Incorrect, the normal supply to both busses 1A and 2A is lost when S/U bus 10 de energizes. Candidates may select this if they do not not recognize that the normal power supply to busses 1A and 2A is lost.

D. Correct, per TM-OP-004, the following conditions are required to be met in order for DG A to ESS 2A breaker 2A20104 to auto close:

• ESS bus voltage <20% for 0.5 sec

• Preferred source breaker 2A201 01 open

• Alternate source breaker 2A20109 open

• 30 cycle time delay

• Bus 2A lockout reset

• DG > 540 rpm, >90% voltage Since the alternate supply is inop, and the normal supply will open when the ESS 2A bus load shed and DG A autostart begins due to the <20% for 0.5 sec Signal, the closure of 2A201 04 will occur after the DG is >540 RPM and >90%

voltage.

Bus 1A alternate supply from S/U bus 20 is still available, and will close in and re-energize bus 1A after bus voltage is lost.

KIA Match Justi'fication:

This question matches the stated KIA since candidates must determine that normal and alternate supplies to bus 2A are lost and recall diesel generator start and load sequences following loss of voltage to bus 2A.

References:

TM-OP-004 rev 2 Reference Required none Learning Objective: 10541.a Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)8 Comments: Created by: T. North, 9/6/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUES1"ION 2 Unit 1 was at 50 % power when a Loss of Offsite Power occurred. After plant conditions have stabilized, the following conditions exist:

• HPCI is in service for RPV pressure control.

• RCIC is in service for RPV level control.

A steam leak then occurs on the HPCI stearn supply line raising temperature in the HPCI pipe routing area to 196 OF in approximately 1 (one) minute, then eventually stabilizing at this temperature.

Which one of the following correctly identifies the status of HPCI AND RCIC 20 minutes later, with NO operator action?

A. HPCI isolated after a 1 second time delay. RCIC will continue to operate indefinitely.

B. HPCI AND RCIC will BOTH continue to operate indefinitely.

C. HPCI AND RCIC BOTH isolated after a 15 minute time delay.

D. HPCI isolated after a 15 minute time delay. RCIC will continue to operate indefinitely.

LOC-23 NRC Exam Rev 3

K&A # 223002 K1 .07 Importance Rating 3.4 QUESTION 2 RO Tier 2 Group 1 K&A Statement: Knowledge of the physical connections and/or cause- effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the Reactor core isolation cooling; Plant-Specific Justification:

A. Incorrect, HPCI isolated after 15 minute TO for a high temp in the pipe routing area, and RCIC will also isolate following a HPClieak in the shared pipe routing area. The 15 minute time delay is applicable only for the pipe routing area isolation signal.

Candidates may select this if they incorrectly recall the isolation setpoint for the shared pipe routing area, and that RCIC will also isolate with a steam leak in the shared area.

B. Incorrect, HPCI and RCIC will both isolate. Candidates may select this if they incorrectly recall the isolation setpoint for the shared pipe routing area, and that RCIC will also isolate with a steam leak in the shared area.

C. Correct, HPCI and RCIC share a common pipe routing area, and the isolation setpoint for both HPCI and RCIC is 167°F after a 15 minute time delay.

Engineering evaluation determined that for a HPClieak in this area, RCIC would also isolate.

O. Incorrect, RCIC will also isolate following a HPCI leak in the shared pipe routing area.

Candidates may select this if they incorrectly recall that RCIC will also isolate with a steam leak in the shared area.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the relationship between Steam leak detection isolation setpoints and the RCtC system.

References:

TM-OP-0598 rev 5 Reference Required none Learning Objective: 2120.b, 2123.0 Question source: SSES OPS_INITIAL_L1CENSE Bank #TMOP059B/2120 004 Question History: Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 55. 41 (b)7 Comments: Created by: Bank Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUES1"ION 3 A fault occurs in 250 VDC Switchgear 1D662, causing the battery charger output breaker to trip and the fuse to the battery to blow.

Which of the following loads would be affected by this event?

A. Main Generator Emergency Seal Oil Pump B. HPCI Auxiliary Oil Pump C. Reactor Feed Pump 1B Emergency Lube Oil Pump D. RCIC Barometric Condenser Vacuum Pump LOC-23 NRC Exam Rev 3

K&A # 263000 K2.01 Importance Rating 3.1 QUESTION 3 RO Tier 2 Group 1 K&A Statement: Knowledge of electrical power supplies to the following: Major D.C. loads Justification:

A. Incorrect, Emergency Seal oil pump is powered from 250vdc control center 1D155 via 1D652. The candidate may select this jf they confuse the ESOP with the main turbine emergency oil pump which is powered by 1D274/1 D662 B. Correct, the HPCI aux oil pump is powered from 10662 via HPCI control center 10274 breaker 31 C. Incorrect, RFP 1B emergency oil pump is powered from 250vdc control center 1D155 via 1D652. The candidate may select this because RFP 1A emergency oil pump is powered from 1D274/1 D662 D. Incorrect, powered by 250 VDC Control Center 1D254 via 1D652. The candidate may select this if they confuse HPCI and RCIC 250 vdc power supplies.

KIA Match Justification:

This question matches the stated KIA since candidates must recall knowledge of power supplies to major 250 VDC oil pumps.

References:

TM-OP-088-FS rev 00, ON-188-001 att Reference Required none B rev 11 Learning Objective: 1383 Question source: SSES OPS_INITIAL_LlCENSE Bank #TMOP088/1393 001 Question History: Bank Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b}7 Comments: Created by:

Reviewed by: T. Ebert, L Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 4 Unit 1 is operating in MODE 2 with ALL Intermediate Range Monitor (lRM) detectors fully inserted into the core.

Which one of the following power supplies, IF LOST, will result in IRM channels A, C, E, and G UNABLE to accurately indicate neutron flux?

A. 120 VAC Panel 1Y216.

B. 120 VAC Pane11Y218.

C. 24 VDC Bus 1D682.

D. 24 VDC Bus 1D672.

LOC-23 NRC Exam Rev 3

K&A # 215003 K2.01 Importance Rating 2.5 QUESTION 4 RO Tier 2 Group 1 K&A Statement: Knowledge of electrical power supplies to the following: IRM chan nels/detectors Justification:

A. Incorrect, 1Y216 does not power any IRM system components. Candidates may select this if they cannot correctly recall power supplies to IRMs.

B. Incorrect, 1Y218 powers the IRM detector drive motors. Since the detectors are fully inserted, loss of this power will not affect the channel's ability to indicate flux.

C. Incorrect, this power supply feeds the div II IRMs, B,D,F,H.

D. Correct, this power supply feeds the div IIRMs, A,C,E,G and its loss will result in these channels being completely de-energized and unable to indicate neutron flux.

KIA Match Justification:

This question matches the stated KIA since candidates must recall power supplies to various IRM detector and channel components and determine the effect on IRM indication.

References:

ON-117-001 rev 30, TM-OP-078B rev Reference Required none 6

Learning Objective: 10230 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)7 Comments: Created by: T. North, 11/14/10 Reviewed by:

LOC-23 NRC Exam Rev 3

QUESTION 5 Unit 1 was operating at full power when a loss of all high pressure feed occurred. The High Pressure Coolant Injection (HPCI) system automatically started and injected due to low RPV water IEwel.

• HPCI is injecting to the RPV at 4000 gpm

• RPV level is +20', steady

• RPV pressure is 934 psig, controlled with turbine bypass valves A logic relay failure causes the HPCI Minimum Flow Isolation valve, HV-155 F012, to fully open.

Determine the impact of this failure on Suppression Pool water level if HPCI were to remain in its current line up:

(consider ONL Y the effect of the HPClline up)

Suppression Pool water level wilL ..

A. REMAIN UNCHANGED because Suppression Pool water will be short cycled back to the Suppression Pool.

B. REMAIN UNCHANGED because Condensate Storage Tank water will be short-cycled back to the Condensate Storage Tank.

C. RISE because Condensate Storage Tank water will be diverted to the Suppression Pool.

D. LOWER because Suppression Pool water will be diverted to the Condensate Storage Tank.

LOC-23 NRC Exam Rev 3

K&A # 206000 K3.03 Importance Rating 3.4 QUESTION 5 RO Tier 2 Group 1 K&A Statement: Knowledge of the effE~ct that a loss or malfunction of the HIGH PRESSURE COOLANT INJECTION SYSTEM will have on following: Suppression pool level control: BWR-2,3,4 Justification:

A. Incorrect, level will rise. Normal suction path is from the CST. Candidates may select this if they incorrectly recall the normal HPCI suction path.

B. Incorrect, level will rise. Min flow line directs water to the SP. Candidates may select this if they incorrectly recall the flow path of the HPCI min flow line.

C. Correct. HPCI normal suction path is from the CST. If the min flow valve were to open with HPCI injecting to the RPV at 4000 gpm, CST water will be diverted to the SP causing SP level to rise.

D. Incorrect, the normal suction path is NOT from the SP, and the min flow line directs water to the SP, NOT the CST. Candidates may select this if they incorrectly recall the normal HPCI suction path and min How line path.

KIA Match Justification:

This question matches the stated KIA since candidates must determine the effect of a failure of the HPCI minimum flow line on suppression pool water level.

References:

TM-OP-052 ST & PG, rev 4 Reference Required none Learning Objective: 2035.d Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 7/30/10 Reviewed by: T. Ebert, L Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 6 Unit 1 is operating at full power when the "A" End of Cycle Recirc Pump Trip (EOC-RPT) logic circuit fails.

• The "A" EOC*RPT trip system is declared inoperable because it is not capable of generating the required trip signal to the recirc system.

• A spurious Main Generator Load Reject scram then occurs.

Which one of the following describes the result of these events?

A. 80TH recirc pumps will trip becausE! the "8" EOC-RPT trip system will trip one RPT breaker for each recirc pump.

8. ONLY the "8" recirc pump will trip because the "8" EOC*RPT trip system trips both RPT breakers for that pump.

C. NEITHER recirc pump will trip because each RPT breaker requires input from 80TH the "A" and "8" EOC-RPT trip systems to function.

D. NEITHER recirc pump will trip because the EOC-RPT function is only activated when less than 26% power.

LOC-23 NRC Exam Rev 3

K&A # 212000 K3.11 Importance Rating 3.0 QUESTION 6 RO Tier 2 Group 1 K&A Statement: Knowledge of the effect that a loss or malfunction of the REACTOR PROTECTION SYSTEM will have on following:

Recirculation system Justification:

A. Correct, each EOC-RPT trip system operates independently by opening an RPT breaker in each recirc pump.

B. Incorrect, both pumps will trip. Candidates may select this if they do not correctly recall the interrelationship between the EOC-RPT logic and recirc pump breakers.

C. Incorrect, both pumps will trip. Candidates may select this if they do not correctly recall the interrelationship between the EOC*RPT logic and recirc pump breakers.

D. Incorrect, the EOC-RPT is activated by the turbine stop valve or load reject scram signals which are active when above 26%. Candidates may select this if they do not correctly recall when EOC-RPT function is active.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the relationship between the EOC-RPT function of RPS, and determine the effect of a failure of that function on the reactor recirc pump breakers.

References:

TM-OP-064C rev 10, TM-OP-058 rev 9 Reference Required none Learning Objective: 2526.aa Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 5/19/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 7 Unit 1 is operating in single loop with the "A" recirc loop in operation.

ALL required actions have been taken for single loop operation Based on these conditions, what is the APRM Simulated Thermal Power (STP)

UPSCALE TRIP setpoint?

A. .55W - 8.7 + 54.2, clamped at 118%

B. .55W - 8.7 + 58.7, clamped at 113.5%

C. .55W + 54.2, clamped at 118%

D. .55W + 58.7, clamped at 113.5%

LOC*23 NRC Exam Rev 3

K&A # 215005 K4.07 Importance Rating 3.7 QUESTION 7 RO Tier 2 Group 1 K&A Statement: Knowledge of AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM design feature(s) and/or interlocks which provide for the following: Flow biased trip setpoints Justification:

A. Incorrect, the setpoint is "+5B.7" and is clamped at 113.5. candidates may select this if they confuse the upscale alarm setpoint with the trip setpoint and do not recall the clamped value.

B. Correct, when in SLO, the normal STP trip setpoint of .55w+5S.7 is modified by subtracting .LlW (S.7)

C. Incorrect, the setpoint is "+58.7", is clamped at 113.5 and is modified for SLO by u_

B. 7". Candidates may select this if they (;onfuse the upscale alarm setpoint with the trip setpoint, do not recall the clamped value or the SLO value.

D. Incorrect, the normal scram setpoint is modified by -!:::"W (B.7) for SLO. Candidates may select this if they do not correctly recall the SLO modifier value.

KIA Match Justification:

This question matches the stated KIA since candidates must recall knowledge of the APRM flow biased trip setpoint for the given flow conditions.

References:

TM-OP-07BD rev 6 Reference Required none Learning Objective: 15710.c Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)5 Comments: Created by: T. North, 5/20/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 8 Unit 1 is shutdown following an automatic scram with the following conditions present:

• RPV pressure is 900 psig, steady

• RPV level is +25 inches, up slow

• RCIC is injecting to the RPV at 600 gpm with the RCIC flow controller in AUTOMATIC The PCOP then adjusts the RCIC flow controller setpoint thumbwheel to 60 gpm.

Which one of the following describes the RCIC system response:

The RCIC Flow Controller will reduce and stabilize ...

A. RCIC RPV INJECTION FLOW at 60 gpm; RCIC minimum flow isolation valve FV-149-F019 will remain CLOSED resulting in ZERO flow to the suppression pool.

B. TOTAL RCIC PUMP FLOW at 60 gpm; RCIC minimum flow isolation valve FV-149-F019 will automatically OPEN resulting in approximately 50 cyo of the total pump now to the suppression pool.

C. TOTAL RCIC PUMP FLOW at 60 gpm; RCIC minimum flow isolation valve FV-149-F019 will automatically OPEN resulting in ALL of the pump lrlow to the suppression pool.

D. RCIC RPV INJECTION FLOW at 60 gpm; RCIC minimum flow isolation valve FV-149-F019 will automatically OPEN resulting in approximately 75 gpm to the suppression pool.

LOC-23 NRC Exam Rev 3

K&A # 217000 K4.03 Importance Rating 2.9 QUESTION 8 RO Tier 2 Group 1 K&A Statement: Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design feature(s) and/or interlocks which provide for the following: Prevents pump over heating Justification:

A. Incorrect, the min flow valve will open when flow drops below 70 gpm. The candidate may select this if they do not recall the correct min flow valve setpoint.

B. Incorrect, the flow controller does not s~nse min flow line flow, and with the controller in auto it will adjust RCIC pump speed to maintain RPV injection flow ONLY at 60 gpm. The min flow line is designed to pass approximately 75 gpm, therefore TOTAL RCIC pump flow will be significantly greater than 60 gpm. The candidate may select this if they do not recall that the flow element does not sense min flow line flow, and that the min flow line will pass 75 gpm by design.

C. Incorrect, the flow controller does not sense min flow line flow, and with the controller in auto it will adjust RCIC pump speed to maintain RPV injection flow ONLY at 60 gpm. The candidate may select this if they do not recall that the flow element does not sense min flow line flow, and they disregard the effect of the min flow line flow restrictors and assume that the min flow line will pass ALL of the 60 gpm.

D. Correct, the min flow valve will automatically open when pump flow is reduced below 70 gpm. The flow controller will continue to maintain RCIC pump speed such that RPV injection flow is stabilized 60 gpm regardless of bypass flow through the min flow line, because the flow element does not sense flow in the min flow line. The min flow line orifices will maintain flow to the supp pool at approximately 75 gpm by design.

KIA Match Justification:

This question matches the stated KIA since the candidates are required to recall facts regarding automatic operation of the min flow isolation valve, the design flowrate of the min flow line. and that the min flow line flow is not sensed by the RCIC flow element. The purpose of establishing flow in this min flow line is to prevent RCIC pump overheating when pump flow is reduced below a certain value.

References:

TM-OP-050 rev 4 Reference Required none Learning Objective: 2008.i, 2018.c, 2012.c Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)5 Comments: Created by: T. North, 8/29/10 Reviewed by: T. Ebert, L.. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 9 During a Unit 1 reactor startup, the ROD OUT BLOCK Annunciator alarms. The following neutron monitoring conditions exist:

• SRMs A and B ..... Fully Inserted, reading 8 x 104 cps.

• SRMs C and D.... Partially withdrawn, reading 60 cps.

• Reactor Period .... +200 seconds.

• IRMs ................. Fully inserted, reading 4 on Range 1.

What one of the following actions will clear tile ROD OUT BLOCK condition and permit continued rod withdrawal?

The ROD OUT BLOCK will clear if ...

A. SRM Detectors C and D are driven in until they indicate greater than 100 cps.

B. a control rod is inserted until count rate is less than 1X1 04 cps.

C. power is allowed to continue to rise until the IRMs indicate above 5 on Range 1.

D. SRM Detectors A and B are driven out until count rate is less than 1X103 cps.

LOC-23 NRC Exam Rev 3

K&A # 215004 K5.01 Importance Rating 2.6 QUESTION 9 RO Tier 2 Group 1 K&A Statement: Knowledge of the operational implications of the following concepts as they apply to SOURCE RANGE MONITOR (SRM)

SYSTEM: Detector operation Justification:

A. Correct. The retract permit rod block is causing the condition.

B. Incorrect. High count rate is 2E5. Candidates may select this if they do not correctly recall SRM rod block setpoints.

C. Incorrect. The IRM downscale is bypassed on Range 1, SRM Retract permit is bypassed @ Range 3. Candidates may select this if they do not correctly recall SRM retract permit logic.

D. Incorrect. High count rate is 2E5. Candidates may select this if they do not correctly recall SRM rod block setpoints.

KIA Match Justification:

This question matches the stated KIA since candidates must recall how SRM detector operation impacts plant operation.

References:

TM-OP-078A rev 3 Reference Required none Learning Objective: 10026 Question source: SSES OPS_I NITIAL_LlCENSE Bank # TMOP078A110026 001 Question History: Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)5 Comments: Created by:

Reviewed by: T. Ebert, L. Casperson 11-12-1 0 LOC-23 NRC Exam Rev 3

QUESTION 10 Unit 1 is operating at full power when the in-service Containment Instrument Gas (CIG) compressor suction filter becomes completely plugged.

This results in a CIG header pressure dropping below 142 psig.

Which one of the following describes the impact of this condition?

A. The standby suction filter will automatically shift into service, but the GIG compressor will trip and control of SRVs, MSIVs, and ADS valves will be lost until the compressor can be restarted.

B. The standby suction filter will automatically shift into service and the GIG compressor will remain in service with NO interruption to GIG loads.

C. The running CIG compressor will trip due to loss of suction, and control of SRVs, MSIVs, and ADS valves will be lost until the standby compressor can be restarted.

D. The running CIG compressor will trip due to loss of suction, but will NOT IMMEDIATELY interrupt control of GIG loads due to the backup gas bottles and accumulators.

LOC-23 NRC Exam Rev 3

K&A # 300000 K5.13 Importance Rating 2.9 QUESTION 10 RO Tier 2 Group 1 K&A Statement: Knowledge of the operational implications of the following concepts as they apply to the INSTRUMENT AIR SYSTEM:

Filters Justification:

A. Incorrect, the standby filter must be manually placed in service. Candidates may select this if they do not correctly recall operational details regarding CIG filter operation.

B. Incorrect, the standby filter must be manually placed in service. Candidates may select this if they do not correctly recall operational details regarding CIG filter operation.

C. Incorrect, control of MSIVs and SRVs will not be immediately impacted. Candidates may select this if they do not correctly evaluate the impact of low CIG pressure resulting from compressor trip.

D. Correct: the compressors will trip on low suction pressure, and the backup gas bottles and accumulator will automatically charge the 150# and 90#

headers respectively, resulting in no immediate loss of component control.

KIA Match Justification:

This question matches the stated KIA since canclidates must determine the effect of CIG filter failure on plant operation.

References:

ON-125-001 rev 12 Reference Required none Learning Objective: 1595.C, 1592.m Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X ComprehensionlAnalysis:

10CFR55 41 (b)5 Comments: Created by: T. North, 5/20/10 Reviewed by: T. Ebert, L Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 11 Unit 1 is in Mode 3; Unit 2 is in Mode 1.

A ground fault occurs on OB516, "DIESEL GEI\IERATOR 'A' ESS 480V MOTOR CONTROL CENTER".

Which one of the following statements describes the effect of the above condition?

A. Unit 1 is NOT affected; Unit 2 Battery Charger 2D613 transfers to its alternate AC source.

B. Unit 2 is NOT affected; Unit 1 Battery Charger 1D613 de-energizes, and battery 1D61 0 will carry the DC loads.

C. Unit 1 AND Unit 2 Battery Chargers 1D613 AND 2D613 are BOTH lost and batteries 1D61 0 and 2D610 will carry the DC loads.

D. Unit 1 AND Unit 2 Battery Chargers 'I D613 AND 2D613 BOTH transfer to their alternate AC source.

LOC-23 NRC Exam Rev 3

K&A # 263000 K6.01 Importance Rating 3.2 QUESTION 11 RO Tier 2 Group 1 K&A Statement: Knowledge of the effect that a loss or malfunction of the following will have on the D.C. ELECTRICAL DISTRIBUTION: A.C.

electrical distribution Justification:

A. Incorrect, Both unit 1 and 2 battery chaq~ers are affected. There is no alternate source of AC power for these battery chargers. Candidates may select this if they do not correctly recall the physical arrangement of chargers 1D613 and 20613.

B. Incorrect, Both unit 1 and 2 battery chargers are affected. Candidates may select this if they do not recall that OB516 supplies both chargers.

C. Correct, OB516 is the AC supply to BOTH unit 1 and 2 battery chargers. With OB516 de-energized, both unit's OC loads supplied by 20610 and 10610 will be supplied by their respective batteries. The batteries are rated for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

D. Incorrect, there is no alternate source of AC power for these battery chargers.

Candidates may select this if they do not correctly recall the physical arrangement of chargers 10613 and 2D613.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the physical connections and relationship between DC battery chargers and their AC power source, and determine the effect of the loss of the AC supply on those battery chargers and the associated DC loads.

References:

TM-OP-002 rev 5; ON-104-201 rev 13 Reference Required none Learning Objective: 1431.a Question source: SSES OPS_INITIAL_LlCENSE Bank # TMOP002/1431 001 Question History: Bank Cognitive level: Memory/Fundamental knowledge: X Comprehension!Analysis:

10CFR55 41 (b)7 Comments: Created by: Bank Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 12 The Unit 1 Core Spray (CS) Initiation Logic Channel A has experienced a loss of power from 125 VDC Class 1E Bus A (1 D614).

If a valid High Drywell pressure condition were to occur, which one of the following describes how CS pumps respond?

A. ONLY Band D pumps start because logic channel B can independently initiate the B CS loop.

B. ONLY A and B pumps start because logic channel B can independently start one pump in each CS loop.

C. ALL 4 pumps start because logic channel B can independently initiate both CS loops.

D. ALL 4 pumps remain OFF because BOTH logic channels are required to initiate each CS loop.

LOC-23 NRC Exam Rev 3

K&A # 209001 K6.04 Importance Rating 2.8 QUESTION 12 RO Tier 2 Group 1 K&A Statement: Knowledge of the effect that a loss or malfunction of the following will have on the LOW PRESSURE CORE SPRAY SYSTEM:

D.C. power Justification:

A. Correct, Only the div 2 pumps start because the B logic has power and can only provide start signals to B loop pumps. Loop A pumps will not start without power to the A initiation logic.

B. Incorrect, Only the loop B pumps receive start signals. Candidates may select this if they do not correctly recall the core spray logic arrangement.

C. Incorrect, see above D. Incorrect, see above KIA Match Justification:

This question matches the stated KIA since candidates must evaluate the failure of DC power to core spray logic and determine the resultant effect following a LOCA signal.

References:

TM-OP-051 ST & PG rev 2; Reference Required none E156-sh 1 rev 21, -sh 2 rev 21, *'sh 3 rev 26, -sh 4 rev 25 Learning Objective: 2093.d Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 7/30/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 13 Unit 1 is in the process of a plant startup after a refueling outage.

• Containment inerting is in progress using "A" Train of Standby Gas Treatment (SGTS)

• "A" SGTS is running and is aligned to take suction from Unit 1 Primary Containment only.

• Oxygen concentration at 20% and going down slowly.

A malfunction in the Unit 1 PCIS logic causes a false RB Zone 3 Isolation Signal on a -38" reactor water level signal to be initiated. The "A" SGTS system responds as designed.

Based on this malfunction, what is the resultant effect on the oxygen concentration in Primary Containment?

Primary Containment oxygen concentration will:

A. RISE due to the increased flow due to the automatic start of the "B" SGTS Train.

B. continue to LOWER but at a slower rate because the Nitrogen supply valves (SV-15767 & SV-15789) remain open.

C. LOWER faster due to the resulting automatic rise in "A" SGTS system flow caused by the zone 3 isolation signal.

D. remain CONSTANT because the primary containment suction dampers (HD17508A & B) closed.

LOC-23 NRC Exam Rev 3

K&A # 261000 A 1 .05 Importance Rating 2.7 QUESTION 13 RO Tier 2 Group 1 K&A Statement: Ability to predict and/or monitor changes in parameters associated with operating the STANDBY GAS TREATMENT SYSTEM controls including: Primary containment oxygen level:

Mark-I&II Justification:

A. Incorrect, 02 levels will remain constant The "B" SGTS train will not start, and this would not cause increased flow from PC. Candidates may select this if they do not correctly recall SGTS response to a -38" signal.

B. Incorrect, SGTS will no longer take suction from PC and the N2 supply valves isolate at -38". Candidates may select this if they do not correctly recall the response to a 38" signal.

C. Incorrect, SGTS flow will not affect 02 levels since PC suction dampers are closed.

Candidates may select this if they do not correctly recall SGTS response to a -38" signal.

D. Correct, the -38" isolation signal will (:ause the PC purge dampers and N2 supply valves to close, resulting in SGTS no longer reducing 02 levels inside PC and no additional N2 being added. 02 levels will remain constant.

KIA Match Justification:

This question matches the stated KIA since candidates must predict changes in oxygen levels following a re-alignment of SGTS during de-inerting activities.

References:

TM-OP-070 rev 5, OP-173-001 rev 37, Reference Required none ON-159-002 att B rev 29 Learning Objective: 1985.j / 1991.a Question source: INPO exam bank # 23535 Question History: Columbia station 2003 NRC exam Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)5 Comments: Modified by: Bank Reviewed by: T_ Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUES1"ION 14 Unit 1 has experienced a LOGA concurrent with a loss of high pressure injection sources. The following conditions are present:

• RPV water level is -135", down slow

• RPV pressure is 475 psig, down slow

• Drywell pressure is 7.5 psig, up slow Which one of the following is true regarding the LPGI system under these conditions?

A.

• All LPGI injection valves are cummtly open;

• Injection will begin when the testable check valves, F050A and B, open at approximately 420 psig.

B.

• LPGI injection valves F015A and B are shut and will open when pressure drops to 420 psig;

• Injection will begin when the testable check valves F050A and B open when RPV pressure reaches LPGI pump shutoff head.

G.

• LPGI injection valves F015A and B are shut;

• Injection will begin when testable check valves, F050A and B, and F015A and B all open when pressure drops to 420 psig.

D.

• All LPGI injection valves are currently open;

• Injection will begin when the testable check valves, F050A and B, open when RPV pressure reaches LPGI pump shutoff head.

LOC-23 NRC Exam Rev 3

K&A # 203000 A 1.02 Importance Rating 3.9 QUESTION 14 RO Tier 2 Group 1 K&A Statement: Ability to predict and/or monitor changes in parameters associated with operating the RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) controls including: Reactor pressure Justification:

A. Incorrect, F015A and B are shut until <420#. F050A and B open when RPV pressure reaches LPCI pump shutoff head (approximately 275-300 psig). Candidates may select this if they do not correctly recall the response of RHR valves during depressurization.

B. Correct, the LPCI auto injection setpoints have been reached, but F015A and B remain closed until <420#. F050A and B remain closed until pressure below LPCI pump shutoff head (approximately 275-300 psig), at which time vessel injection will begin.

C. Incorrect, F050A and B will not open until RPV pressure reaches LPCI pump shutoff head. Candidates may select this if they do not correctly recall the response of RHR valves during depressurization.

D. Incorrect, F015A and B are shut until <420#. Candidates may select this if they do not correctly recall the response of RHR valves during depressurization.

KIA Match Justification:

This question matches the stated KIA since candidates must be able to predict the response of LPCI injection vavles during RPV depressurization while in the LPCI injection mode.

References:

TM-OP-050 rev 4 Reference Required none Learning Objective: 196.0 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)7 Comments: Created by: T. North, 5/20/10 Reviewed by: T. Ebert, L Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 15 Unit 2 is operating in mode 1 when power is lost to 480 VAC Panel 2B236.

An ATWS occurs and the Unit Supervisor directs the PCOP to inject Standby Liquid Control (SBLC).

Which one of the following is correct concerning boron injection?

A. ONLY SBLC Pump A is available.

Initiate SBLC per OP*253*001, "Standby Liquid Control System".

B. ONLY SBLC Pump B is available.

Initiate SBLC per OP*253*001, "Standby Liquid Control System".

C. BOTH SBLC subsystems are available.

Initiate SBLC per OP-253*001, "Standby Liquid Control System".

D. NEITHER SBLC subsystem is available.

Implement ES-250-002, "Boron Injection Via RCIC".

LOC-23 NRC Exam Rev 3

K&A # 211000 A2.03 Importance Rating 3.2 QUESTION 15 RO Tier 2 Group 1 K&A Statement: Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: A.C.

power failures Justification:

A. Incorrect, SBLC pump A is powered from 2B236, therefore it will not start if its handswitch is placed in start. Candidates may select this if they do not recall the 480 vac power supply to Unit 2's SLC Pumps.

B. Correct, SBlC pump B (2B217) and squib valve A (2Y216) are available for boron injection. Note Unit 2 squib valve power supplies are the reverse of Unit 1 (A from Y236, B from Y216)

C. Incorrect, 2B236 is the power supply for A SBLC Pump and 2Y236, which is the power supply to B Squib valve. Therefore, only the B SBLC Pump can inject via the A Squib valve. Candidates may select this if they do not recall the 480/120 vac power supply to Unit 2's SLC Pumps and squib valves.

D. Incorrect, as stated above, B SBLC Pump is able to inject. Although implementing ES-150-002 would allow the use of RCIC to inject boron, it is only implemented in the event that SBLC can not inject boron. Candidates may select this if they do not recall the 480/120 vac power supply to Unit 2's SLC Pumps and squib valves.

KIA Match Justification:

This question matches the stated KIA since candidates must determine the impact on SBLC due to the loss of 480 vac and 120vac panels powering some SBLC components, AND determine the correct method and procedure required to inject boron with this AC power unavailable.

References:

TM-OP-053-ST rev 9, TM-OP-053-FS Reference Required none rev 2, ON-217-001 Att. H, rev 24 Learning Objective: 1214.c Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)5 Comments: Created by: M. Jacopetti, 11130/10 Reviewed by: M. Shaffer, G. Shellenberger, R. Klinefelter 11-30-10 LOC-23 NRC Exam Rev 3

QUES1"ION 16 Unit 1 is operating at full power with the following conditions present:

• 1 Feedwater Flow input is BYPASSED due to deviation >0.50%

• LlC-C32-1 R600, FW Level Ctl/Demand Signal Controller, is in AUTOMATIC

• RPV level control is selected to 3 ELEMENT

• RPV level stable at +35" A SECOND Feedwater Flow input then FAILS DOWNSCALE.

Which one of the following describes:

(1) The RPV level response (if any), AND; (2) The action(s) that should be taken in accordance with ON-145-001, "RPV Level Control System Malfunction"?

A. (1) RPV level rises to approximately +42", then stabilizes at approximately +35".

(2) Shift ICS/DCS to SINGLE ELEMENT, then manually bypass the second failed RFP flow input.

B. (1) RPV level remains stable at approximately +35".

(2) Maintain ICS/DCS in its current configuration and verify that ICS/DCS automatically bypassed the second failed RFP flow input.

C. (1) RPV level remains stable at approximately +35".

(2) Verify that ICS/DCS automatically shifted to SINGLE ELEMENT control, then manually bypass thH second failed RFP flow input.

D. (1) RPV level rises to approximately +42", then stabilizes several inches ABOVE +35".

(2) Establish manual control of Reactor Feed Pump speeds to return RPV level to +35" until one feedwater flow input can be returned to service.

LOC-23 NRC Exam Rev 3

K&A# 259002 A2.02 Importance Rating 3.3 QUESTION 16 RO Tier 2 Group 1 K&A Statement: Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM; and (b) based on those predictions. use procodures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of any number of reactor feedwater flow inputs Justification:

A. Correct: Per ON-145-001, the 2nd failed feed flow transmitter remains in service until manually bypassed. The downscale signal causes an RPV level rise to 42" in response to a potential reduction in level due to lowering total feed flow signal. Level control remains in 3 element, and restore level to 35".

The 2nd failed flow instrument must be manually bypassed, and ICS/DCS should be shifted to single element.

B. Incorrect, the 2 nd failed instrument remains in service and creates a rising level trainsient; ON-145-001 should be used to transfer level control to single element and manually bypass the 2 nd failed feed flow input. Candidates may select this if they incorrectly assume that ICS/DCS will automatically bypass the second failed flow input, RPV level will remain stable, and that level control should remain in 3 element control.

C. Incorrect, the 2 nd failed instrument remains in service and creates a rising level trainsient. Candidates may select this if they assume that ICS/DCS will automatically shift to single element control and level will remain stable.

D. Incorrect, Level will return to approximately +35". Level control should be shifted to single element control and the second foed flow input must be manually bypassed.

Candidates may select this if they incorrectly assume that ICS/DCS compensate for the lost feed flow input by establishing a higher level setpoint.

KIA Match Justification:

This question matches the stated KIA since candidates must predict plant response to a 2 nd failed feed flow instrument and determine the actions required per the stated procedure.

References:

ON-145-001 rev 27 Reference Required none Learning Objective: 16014.d Question source: New Question History: New Cognitive level: MemorylFundamental knowledge:

ComprehensionlAnalysis: X 10CFR55 41 (b)5 Comments: Created by: T. North, 912110 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 17 Unit 1 has experienced a LOCA resulting in the following:

• RPV level is -160", down fast

• RPV pressure is 40 psig, down slow The crew is directed by EOPs to open 6 ADS valves.

The PCOP manually actuates the ADS logic, then notes the following conditions:

• RPV pressure is 35 psig, down slow

• All 6 ADS valve solenoid energized lights are LIT

• All acoustic monitor LEDs are NOT LIT The PCOP should report that. ..

A. Six ADS valves are open because the acoustic monitor LEDs are normally lit and extinguish when the valves open.

B. Six ADS valves are open because the acoustic monitor may not indicate at low RPV pressures.

C. NO ADS valves are open because the acoustic monitor LEDs will illuminate when the valves open under these conditions.

D. NO ADS valves are open because the ADS valve solenoids de-energize to open the valves.

LOC-23 NRC Exam Rev 3

K&A# 239002 A3.04 Importance Rating 3.6 QUESTION 17 RO Tier 2 Group 1 K&A Statement: Ability to monitor automatic operations of the RELIEF/SAFETY VALVES including: Acoustical monitor noise: Plant-Specific Justification:

A. Incorrect, the LEOs are normally off and are lit when valves open if RPV pressure is suffiently high. Candidates may select this if they do not correctly recall the operation of the acoustic monitor LEOs.

B. Correct, at low RPV pressure (ie, when RD is complete) acoustic monitors may not indicate SRV position. EO-OO()"112 basis states that solenoid energized lights and RPV pressure not rising is sufficient to determine ADS valves open.

C. Incorrect, insufficient indication is available to determine that six ADS valves are open. Candidates may select this if they do not correctly recall the operation of the acoustic monitor LEOs.

D. Incorrect, ADS solenoids energize to open the valves and combined with RPV pressure not rising is sufficient to determine ADS valves open. Candidates may select this if they do not correctly recall the operation of the ADS solenoid valves.

KIA Match Justification:

This question matches the stated KIA since candidates must recall facts regarding acoustic monitors in order to accurately monitor ADS valve operation.

References:

EO-000-112 rev 5; TM-OP-083, rev 8 Reference Required none Learning Objective: 14596 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41(b)7 Comments: Created by: T. North, 7/31/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 18 Unit 1 was operating at full power with all systems in a normal full power lineup when a transient occurs, resulting in drywell pressure rising to 2.0 psig.

In this condition, operators should closely monitor (1) because (2)

A. (1) Reactor Recirc Pump (RRP) motor winding temperatures (2) Reactor Building Closed Cooling Water flow to the RRP motor winding coolers automatically isolated B. (1) Reactor Recirc Pump (RRP) motor winding temperatures (2) RRP motor winding cooling automatically shifted from Reactor Building Chilled Water to Reactor Building Closed Cooling Water C. (1) Reactor Recirc Pump (RRP) motor bearing and seal temperatures (2) Reactor Building Closed Cooling Water flow to the Recirc pump bearing and seal coolers automatically isolated D. (1) Reactor Recirc Pump (RRP) motor bearing and seal temperatures (2) RRP bearing and seal cooling automatically shifted from Reactor Building Chilled Water to Reactor Building Closed Cooling Water LOC-23 NRC Exam Rev 3

K&A # 400000 A3.01 Importance Rating 3.0 QUESTION 18 RO ner 2 Group 1 K&A Statement: Ability to monitor automatic operations of the CCWS including:

Setpoints on instrument signal levels for normal operations, warnings, and trips that are applicable to the CCWS Justification:

A. Incorrect, although cooling flow to the winding coolers will have isolated, the isolation of RBCCW will not affect this load Since it is normally cooled by RBCW. Candidates may select this if they do not recall that the normal cooling supply to the winding coolers is RBCW vice RBCCW.

B. Incorrect, cooling flow to the winding coolers will only shift to RBCCW upon a failure of RBCW. In this condition, RBCW flow to the winding coolers is isolated by the high OW pressure. Candidates may select this if they incorrectly apply the RBCCW shift for these plant conditions.

C. Correct, RBCCW piping that penetrates the primary containment will isolate when OW pressure exceeds 1.72 psig, stopping flow to recirc pump coolers, therefore recirc pump temperatures should be closely monitored.

O. Incorrect, bearing and seal cooler cooling flow is normally provided by RBCCW and will isolate on high OW pressure, therefore there will be no shift in cooling 'flow.

Candidates may select this if they do not correctly recall the normal cooling water supply for these coolers and that cooling flow will isolate on high DW pressure.

KIA Match Justification:

This question matches the stated KIA since the candidates are required to recall the automatic actions associated with the RBCCW and RBCW system when drywell pressure is above the high drywell pressure setpoint, determine which loads are affected by the RBCCW containment isolation and will require additional monitoring.

References:

ON-159-002 rev 29, TM-OP-014 rev 1, Reference Required none TM-OP-064C rev 10, ON-114-001 rev 21 Learning Objective: 1694.c Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)7 Comments: Created by: T. North, 5/2:4/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 19 Unit 1 has experienced a transient that requires rapid depressurization due to a PRIMARY system leaking into SECONDARY containment.

The Automatic Depressurization System (ADS) has been manually initiated by arming and depressing the ADS manual initiation push buttons on 1C601.

Immediately after all ADS valves open, the PCOP reports a very rapid rise in suppression chamber and drywell pressure.

Which one of the following completes the statement below describing the response observed by the PCOP?

This pressure rise is (1) ,and is the result of (2)

A. (1) NORMAL (2) the rapid rush of steam entering the suppression pool below the water line.

B. (1) NORMAL (2) the automatic opening of drywell to suppression pool vacuum breakers.

C. (1) ABNORMAL (2) a failure of primary containment downcomer piping.

D. (1) ABNORMAL (2) a failure of a Safety Relief Valve tailpipe above the water line.

LOC-23 NRC Exam Rev 3

K&A# 218000 A4.01 Importance Rating 4.4 QUESTION 19 RO Tier 2 Group 1 K&A Statement: Ability to manually operate and/or monitor in the control room:

ADS valves Justification:

A. Incorrect, the suppression pool water inventory is sized to accommodate the full energy released by the RPV when ADS is initiated. By design this results in relatively little impact on SP and DW pressure. Candidates may select this if they are unfamiliar with or cannot correctly recall the normal containment pressure response following multiple SRV opening.

B. Incorrect, the DW-SP vacuum breakers are not expected to operate during ADS valve actuation, therefore this pressure rise is considered abnormal. Candidates may select this if they are unfamiliar with the normal containment pressure response following multiple SRV opening.

C. Incorrect, the response is abnormal. A failure of a downcomer would tend to equalize SP and DW pressures, but provided the ADS valves discharge under the water line, this failure would not result in a rapid PC pressure rise. Candidates may select this if they confuse primary containment downcomers with SRV tailpipes or are generally unfamiliar with primary containment structure and design, and potential abnormalities associated with SRV tailpipe failures.

D. Correct, this failure would result in steam energy from the RPV released directly into the suppression chamber air space. This will result in a very rapid rise in SP and DW pressures.

KIA Match Justification:

This question matches the stated KIA since candidates must evaluate the response of parameters that should be monitored after ADS actuation and determine if proper system and component operation has occurred.

References:

TM-OP-059 rev 1 Reference Required none Learning Objective: 2092.d Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 9/1/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUES1"ION 20 Unit 1 is in mode 4, with RHR pump A in shutdown cooling mode.

A LOCA occurs, resulting in RPV level rapidly lowering until the leak is stopped.

RPV level stabilizes at +5" Which one of the following describes the actions the PCOP should take with regard to the RHR system?

A.

• Verify that RHR Shutdown Cooling suction isolation valves F008, F009 AND RHR loop A injection valve F015A automatically SHUT;

• Verify RHR pump A automatically trips.

B.

• Verify that RHR Shutdown Cooling suction isolation valves F008 and F009, AND RHR pump A Shutdown Cooling suction valve F006A automatically SHUT;

• Verify RHR pump A automatically trips.

C.

• Verify that RHR Shutdown Cooling suction isolation valves F008 ,

F009 AND RHR loop A injection valve F015A automatically SHUT;

• Manually trip RHR pump A.

D.

• Verify that RHR Shutdown Cooling suction isolation valves F008 and F009, AND RHR pump A Shutdown Cooling suction valve F006A automatically SHUT;

• Manually trip RHR pump A.

LOC-23 NRC Exam Rev 3

K&A # 205000 A4.02 Importance Rating 3.6 QUES'nON 20 RO Tier 2 Group 1 K&A Statement: Ability to manually operate and/or monitor in the control room:

SDC/RHR suction valves Justification:

A. Correct, when RPV level reaches +13", SDC suction valves F008 & F009 auto close. This results in a loss of suction path automatic trip of RHR pump A, since F004A will also be shut due to the SDC lineup.

S. Incorrect, F006A will remain open. The candidate may select this if they incorrectly believe that F006A will also auto close.

C. Incorrect, the A RHR pump will automatically trip due to loss of suction. The candidate may select this if they incorrectly believe that the RHR pump will remain running with no suction path.

D. Incorrect, F006A will not shut, and RHR pump A will trip. The candidate may select this if they incorrectly believe that the RHR pump will remain running with no suction path and F006A will auto close.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the RHR suction valve interlocks in order to properly monitor the system response following a low RPV level condition.

References:

TM-OP-049 rev 7 Reference Required none Learning Objective: 181.q Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created/Modified by: T. North, 9/2/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 21 EO~100-113, "Level/Power Control", directs operators to reset the Main Generator lockout relays if a Main Turbine trip has occurred.

This action is required to ...

A. maintain normal power to the auxiliary busses by preventing transfer to the startup supply.

B. maintain the MSIVs open by preventing a spurious isolation signal due to loss of auxiliary busses.

C. prevent loss of significant equipment due to an undesired load shed on auxiliary busses.

D. prevent loss of equipment required for accident mitigation due to a load shed of the ESS busses.

LOC*23 NRC Exam Rev 3

K&A# 262001 2.4.18 Importance Rating 3.3 QUESTION 21 RO Tier 2 Group 1 K&A Statement: Emergency Procedures / Plan: Knowledge of the specific bases for EOPs.

Justification:

A. Incorrect, this transfer will not be prevented by resetting the lockouts.

B. Incorrect, a spurious MSIV isolation would not be caused by a loss of the aux busses.

C. Correct, per EO-OOO-113 bases, the intent of this step is to prevent an inappropriate aux bus load shed. Since spurious operation of the load shed will cause an undesirable loss of loads such as condensate pumps, SW pumps, etc, main generator lockouts are reset if RPV water level can be maintained >-129" D. Incorrect, the ESS load shed is desirable in EOPs since it ensures equipment needed for accident mitigation has a reliable source of power.

KIA Match justification:

This question matches the stated KIA since candidates must recall knowledge of EOP action bases related to AC distribution.

References:

EO-000-113 rev 8, SSES-EPG rev 8 Reference Required none Learning Objective: 14613 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)

Comments Created by: T. North, 10/10/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 22 Unit 1 is operating in MODE 1.

During functional testing of the Automatic Depressurization System (ADS) system, it is determined that the "ADS MANUAL 1f\IITIATION A" Pushbutton, HS-B21-1 S30 A, has failed.

The pushbutton CANNOT be armed OR depressed.

Determine the affect this failure will have on the ADS system; AND, given Tech Specs 3.3.5.1, "Emergency Core Cooling System Instrumentation" determine the status of LCO 3.3.5.1 :

A.

• ADS CAN be manually initiated because BOTH the Band D pushbuttons are still operable.

• LCO 3.3.5.1 is NOT met because the required number of manual channels are NOT operable.

B.

• ADS CAN be manually initiated because BOTH the Band D pushbuttons are still operable.

• LCO 3.3.5.1 IS met because the required number of manual channels ARE operable.

C.

• ADS CANNOT be manually initiated because ALL FOUR pushbuttons are required to actuate the logic.

• LCO 3.3.5.1 is NOT met becausE~ the required number of manual channels are NOT operable.

D.

• ADS CANNOT be manually initiated because ALL FOUR pushbuttons are required to actuate the logic.

• LCO 3.3.5.1 IS met because the manual initiation function is NOT required in this condition.

LOC*23 NRC Exam Rev 3

K&A# 2180002.2.40 Importance Rating 3.6 QUESTION 22 RO Tier 2 Group 1 K&A Statement: Equipment Control: Ability to apply technical specifications for a system.

Justification:

A. Correct, arming and depressing the Band D pushbuttons will result in manual initiation of the ADS logic. The minimum number of required channels for the manual function of ADS instrumentation not met per table 3.3.5.1-1, therefore LCO 3.3.5.1 is not met.

B. Incorrect, see A above. Candidates may select this if they do not correctly apply table 3.3.5.1-1 requirements.

C. Incorrect, arming and depressing the Band D pushbuttons will result in manual initiation of the ADS logic. BOTH buttons in EITHER logic can initiate ADS.

Candidates may select this if they incorrectly recall that all 4 buttons are required.

D. Incorrect, ADS can be initiated using the Band D pushbuttons. LCO 3.3.5.1 is not met due to the manual function. Candidates may select this if they incorrectly determine that the manual function is not required in mode 1.

KIA Match Justification:

This question matches the stated KIA since candidates must apply tech specs for the given conditions.

References:

TS 3.3.5.1 rev 3 Reference Required TS 3.3.5.1 Learning Objective: 12701 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)10 Comments: Created by: T. North, 11-12-10 Reviewed by: T. Ebert, L. Casperson 11-12-1 0 LOC-23 NRC Exam Rev 3

QUESTION 23 Which one of the following power supplies is required in order for the High Pressure Coolant System (HPCI) to automatically perform its intended safety function WITHOUT manual operator action?

A. 120 VAC panel1Y216 B. 480 VAC MCC 1B237 C. 125 VDC panel 1D614 D. 125 VOC panel 10624 LOC-23 NRC Exam Rev 3

K&A# 206000 K2.03 Importance Rating 2.8 QUESTION 23 RO Tier 2 Group 1 K&A Statement: Knowledge of electrical power supplies to the following: Initiation logic: BWR-2,3,4 Justification:

A. Incorrect, this panel only provides power to the div 1 steam leak detection, F100 warmup isolation solenoid, and BIS indication. Loss of power to these components would not prevent HPCI auto start on a LOCA signal. Candidates may select this if they do not correctly recall the HPCI initiation logic power supply.

B. Incorrect, this MCC only provides power to F002 inboard isolation valve MOV. Loss of power to this valve would not prevent HPCI auto start on a LOCA signal, since the valve is normally open and would not reposition. Candidates may select this if they do not correctly recall the HPCI initiation logic power supply.

C. Incorrect, this panel only provides power to the div 1 isolation logic; F028, steam line drain solenoid; and F026 barometric condenser drain solenoid. The isolation logic is not required for auto start on a LOCA signal. F026 and F028 would remain closed but are not required for automatic injection on a LOCA signal. Candidates may select this if they do not correctly recall the HPCI initiation logic power supply.

D. Correct, HPCI auto initiation logic is powered by 1D624. Without power to this panel, HPCI would not automatically start and inject when required by a valid LOCA signal.

KIA Match Justification:

This question matches the stated KIA since candidates must correctly recall the power supply for HPCI initiation logic.

References:

TM-OP-052 rev 4 Reference Required none Learning Objective: 10367.c Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 6/2:0/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 24 With both units operating in MODE 1, HSS-0653B, one of the four Channel ICI Common Load Manual Transfer Switches (for diesel generator ESW Valve control and indication and, Diesel Generator Fuel Oil Booster Pump control), is transferred from its NORMAL position to its ALTERNATE position.

Which one of the following statements describes what will occur as a result of this?

The loads powered via HSS-0653B are now powered from (1)  ; AND there will be (2)

A. Unit 2; a MOMENTARY loss of power to the affected loads since this switch is "break-before-make" B. Unit 1; a MOMENTARY loss of power to the affected loads since this switch is "break-before-make" C. Unit 2; NO loss of power to the affected loads since this switch is "make-before break" D. Unit 1; NO loss of power to the affected loads since this switch is "make-before break" LOC-23 NRC Exam Rev 3

K&A # 263000 K4.02 Importance Rating 3.1 QUESTION 24 RO Tier 2 Group 1 K&A Statement: Knowledge of D.C. ELECTRICAL DISTRIBUTION design feature{s) and/or interlocks which provide for the following:

Breaker interlocks, permissives, bypasses and cross ties: Plant Specific Justification:

A. Corrrect, the alternate supply is from unit 2. The switch is break-before-make and will result in a momentary power loss to affected loads B. Incorrect, power supply is from unit 1 in NORMAL. Candidates may select this if they do not correctly recall the normal and alternate power sources for the common DC loads C. Incorrect, power supply is from unit 1 in NORMAL. A momentary power loss to affected loads will occur due to the "break-before-make" switch operation.

Candidates may select this if they do not correctly recall how the transfer switch is interconnected between units and the annunciator arrangement.

D. Incorrect, a momentary power loss to affected loads will occur due to the "break before-make" switch operation. Candidates may select this if they believe this transfer is annunciated.

KIA Match Justification:

This question matches the stated klA since candidates must recall knowledge of DC distribution system cross tie switches.

References:

TM-OP-002-ST rev 5, OP-102-002 rev Reference Required none 13 Learning Objective: 10144 Question source: MODIFIED SSES OPS_INITIAL_L1CENSE Bank

1. TMOP002/1 0144 004 Question History: MODIFIED Bank Cognitive level: Memory/Fundamental knowledge: X ComprehensionlAnalysis:

10CFR55 41{b)7 Comments: modified Modified by: T. North, 11-22-10 to remove reference Reviewed by: E. Brice, A. Avery, 11-23-1 0 to annunciation and add switch details.

LOC-23 NRC Exam Rev 3

QUESTION 25 A failure of the Reactor Protection System (RPS) has occurred at Unit 1 requiring manual initiation of Alternate Rod Insertion (ARI).

The PCOP attempts to arm and depress BOTH the Div 1 AND Div 2 ARI manual initiation pushbuttons on 1C601.

The DIV 1 pushbutton is armed and depressed successfully; HOWEVER, the arming collar on the DIV 2 ARI manual pushbutton FAILS; AND the DIV 2 pushbutton CANNOT be armed OH depressed.

The PCOM mustthen report that (1) ,because (2) ARI vent and block valves repositioned.

A. (1) ALL Control Rods inserted (2) ALL FOUR B. (1) Control Rods DID NOT insert (2) ONLY the DIV 1 C. (1) Control Rods DID NOT insert (2) NONE of the D. (1) ALL Control Rods inserted (2) ONLY the DIV 1 LOC-23 NRC Exam Rev 3

K&A # 212000 A4.16 Importance Rating 4.4 QUESTION 25 RO Tier 2 Group 1 K&A Statement: Ability to manually operate and/or monitor in the control room:

Manually activate anticipated transient without SCRAM circuitry/RRCS: Plant*Specific Justification:

A. Incorrect, the div 2 ARI valves can only be repositioned by the div 2 logic. 80th div 1 and div 2 ARI valves are required to cause rod motion. Candidates may select this if they fail to recall that ARI logic requires both div 1 and div 2 pushbuttons to satisfy the complete logic and reposition all 4 vcllves.

B. Correct, the failure of div 2 logic results in only the div 1 valves repositioning, and since all 4 ARI valves are required to reposition to cause rod motion, no rod motion occurs.

C. Incorrect, the div 1 valves will reposition. Candidates may select this if they incorrectly believe that both pushbuttons are required to actuate either division's logic.

D. Incorrect, no rod motion will occur. Candidates may select this if they fail to recall that the scram air header requires both cliv 1 and div 2 valves to vent and cause rod motion.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the logic arrangement of the ARI system to predict what the correct report will be following this failure.

References:

TM-OP*058 rev 9 Reference Required none Learning Objective: 11480.j Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 9/18/10 Reviewed by: T. Ebert, L. Casperson 11*12*10 LOC-23 NRC Exam Rev 3

QUESTION 26 Unit 1 was operating at 20 percent power when a loss of all high pressure feed occurs.

• Reactor water level dropped to -140 inches

• Offsite power is available

• Unit 2 RHR pumps are in standby Which one of the following describes the Unit 1 RHR Pump start sequence under these conditions?

A. All four RHR Pumps start immediately.

B. A and B RHR Pumps start immediately; C and D RHR pumps start after a 7 second time delay.

C. A and B RHR Pumps start after a 3 second time delay; C and D RHR pumps start after a 7 second time delay.

D. All four RHR Pumps start after a 3 second time delay.

LOC*23 NRC Exam Rev 3

K&A # 203000 A3.08 Importance Rating 4.1 QUESTION 26 RO Tier 2 Group 1 K&A Statement: Ability to monitor aut()matic operations of the RHR/LPCI:

INJECTION MODE (PLANT SPECIFIC) including: System initiation sequence Justification:

A. Incorrect, C/D on seven second delay. This answer would be correct only if all 4 ESS busses were energized by the DGs. Candidates may select this if they do not recall that C/D have a 7 sec time delay.

B. Correct, AlB start immediately, C/O on 7 sec TO to prevent OIL the ESS busses C. Incorrect, AlB start immediately, 3 sec TD is only applicable with LOOP. Candidates may select this if the incorrectly apply the 3 sec LOOP time delay to this plant condition.

D. Incorrect, 3 sec TD is only applicable with LOOP. Candidates may select this if the incorrectly apply the 3 sec LOOP time delay to this plant condition.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the correct LPCI pump start sequence to properly monitor automatic system operation.

References:

TM-OP-049 rev 7 Reference Required none Learning Objective: 181.1 Question source: SSES OPS_INITIAL_LlCENSE Bank #TMOP049/192 001 Question History: Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: Bank Reviewed by: T. Ebert, L Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 27 A LOCA has occurred on Unit 1 causing drywell pressure and temperature to rise.

The Unit Supervisor has determined that drywell sprays are REQUIRED in accordance with EO-1 00-1 03, "PC Control", HOWEVER, PCOs report that NO Residual Heat Removal (RHR) Pumps are operable.

Which one of the following describes ALTERNATE equipment that can be used to perform drywell sprays?

A. BOTH Emergency Service Water, AND RHR Service Water pumps.

B. ONLY RHR Service Water pumps.

C. ONLY Emergency Service Water pumps.

D. Emergency Service Water, RHR Service Water OR Fire pumps.

LOC-23 NRC Exam Rev 3

K&A # 226001 K1.11 Importance Rating 2.8 QUESTION 27 RO Tier 2 Group 2 K&A Statement: Knowledge of the physical connections and/or cause- effect relationships between RHA/LPCI: CONTAINMENT SPRAY SYSTEM MODE and the following: Component cooling water systems Justification:

A. Incorrect, only RHRSW pumps can be aligned for this purpose.

B. Correct, RHRSW pumps can be aligned to provide spray flow by opening RHRlRHRSW crosstie valves per PO-116-001.

C. Incorrect, see A above. Candidates may select this if they do not correctly recall the physical connections between ESW and RHR.

D. Incorrect, see A above. Candidates may select this if they do not correctly recall the physical connections between ESW, Fire suppression and RHA.

KIA Match Justification:

This question matches the stated KIA since candidates are required to recall the physical connections and relationship between RHR and RHRSW.

References:

OP-116-001 rev 29, EO-000-103 rev 7 Reference Required none Learning Objective: 14621 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)8 Comments: Created by: T. North, 11/22110 Reviewed by: E. Brice, A. Avery, 11-23-10 LOC*23 NRC Exam Rev 3

QUESTION 28 Unit 1 was operating at 90% power when the fOllow~~uence of events occur:

The inboard MSIVs failed closed. "

RPV level reached -42" before being restored to 35".

• RPV pressure dropped to 1,020 psig and is being maintained between 800 and 1,087 psig.

• RHR Loop A was placed in Suppression Pool Cooling.

• The feeder breaker to MCC 1B237 tripped

• A small LOCA caused Drywell pressure to reach 2.5 psig.

Which valve(s) will reposition as a result of these conditions?

A. HV-151-F007A, RHR Pump Ale Min Flow Valve.

B. HV-151~F017A, RHR Injection Flow Control Valve, AND HV-151-F028A, Suppression Spray Test Shutoff Valve.

C. HV-151-F027A, Suppression Pool Spray Control Valve, AND HV-151 F048A, Heat Exchanger A Shell Side Bypass Valve.

D. HV-11210A, RHR Service Water Heat Exchanger Inlet Valve.

LOC-23 NRC Exam Rev 3

K&A # 219000 K2.01 Importance Rating 2.5 QUESTION 28 RO Tier 2 Group 2 K&A Statement: RHA/LPCI: Torus/Suppression Pool Cooling Mode:

Knowledge of electrical power supplies to the following: Valves Justification:

A. Correct, HV-151-F007A (powered from 18219) will receive an open signal due to low system flow when HV-151-F028A (powered from 18216) closes as a result of Drywell pressure exceeding 1,,72 psig.

B. Incorrect, HV-151-F017A is closed per OP-149-004 when aligning for Suppression Pool Cooling and will not automatically reposition unless RPV pressure reaches 420 psig. HV-151-F028A (powered from 1B216) is open while in supp pool cooling and does close due to the high Drywell pressure signal.

C. Incorrect, HV-1S1-048A (powered from 18237) will receive an open signal from the high Drywell pressure but will not reposition due to the loss of power. HV-151 F027A (powered from 1B236) is already closed when it receives a closed signal due to the high Drywell pressure.

D. Incorrect, HV-11210A heat exchanger RHRSW inlet valve, although powered from MCC 1B237, does not receive a closed signal from the -38" RPV level signal and will remain open. It is the RHRSW Pump that trips on the -38" signal.

KIA Match Justification:

This question matches the stated KIA since candidates are required to correctly recall the power supplies to various motor operated valves utilized in suppression pool cooling mode of RHR.

References:

TM-OP-049 rev 7 Reference Required none Learning Objective: 10499.b Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)7 Comments: Created by: M. Jacopetti 11/30/10 Reviewed by: M. Shaffer, G. Shellenberger, R. Klinefelter 11-30-10 LOC*23 NRC Exam Rev 3

QUESTION 29 Unit 1 is operating in mode 1 with the following steady state conditions present:

• Reactor power: 90%

• Recirc flow: 89% of rated

• RPV pressure: 1015 psig A failure of the Reactor Recirc Flow Control System then causes a slow RISE in the speed of BOTH Recirc pumps.

Predict the INITIAL effect this failure will have on the Main Turbine control system, and RPV pressure:

A. Main Turbine Bypass Valves will slowly open as reactor power rises above 90%, and RPV pressure will be maintained at 1015 pSig.

B. Main Turbine Bypass Valves will slowly open as reactor power rises above 90%, and RPV pressure will slowly rise above 1015 psig.

C. Main Turbine Control Valves will slowly open as reactor power rises above 90%, and RPV pressure will be maintained at 1015 psig.

D. Main Turbine Control Valves will slowly open as reactor power rises above 90%, and RPV pressure will slowly rise above 1015 psig.

LOC-23 NRC Exam Rev 3

K&A# 202002 K3.04 Importance Rating 2.9 QUESTION 29 RO Tier 2 Group 2 K&A Statement: Knowledge of the effect that a loss or malfunction of the RECIRCULATION FLOW CONTROL SYSTEM will have on following: Reactor/turbine pressure regulation system Justification:

A. Incorrect, bypass valves will not begin to open until turbine load limits are reached.

Pressure will rise as reactor power rises due to the rise in recirc pump speed.

Candidates may select this if they do not correctly recall that the load limit is set above 100% turbine load.

B. Incorrect, bypass valves will not begin to open until turbine load limits are reached.

Pressure will rise as reactor power rises due to the rise in recirc pump speed.

Candidates may select this if they do not correctly recall that the load limit is set above 100% turbine load.

C. Incorrect, pressure will rise as reactor power rises due to the rise in recirc pump speed. Candidates may select this if they do not recall that RPV pressure will rise as power is ramped up.

D. Correct, the rise in recirc pump speed will cause a corresponding rise in reactor power. The turbine control system will allow RPV pressure to rise proportionally with the power rise and cause the control valves to open, increasing turbine load.

KIA Match Justification:

This question matches the stated KIA since candidates must predict the response of the main turbine control system following a failure of the reactor recirc flow control system.

References:

TM-OP-093L rev 6, TM-OP-064E rev 0 Reference Required none Learning Objective: 10341.a Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 9/19/101 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 30 Input to the Reactor Feed Pump level 8 trip logic is provided by Reactor Vessel Narrow Range Water Level Transmitters A, Band C.

With Unit 1 at full power:

PREDICT the impact on Reactor Feed Pumps (RFP) if the "A" Narrow Range Water Level Transmitter were to fail UPSCALE (false high RPV level signa!),

AI\ID why:

A. All three (3) RFPs will trip because the logic only requires one (1) upscale signal to cause a trip.

B. ONLY RFP "A" will trip because each transmitter only feeds its respective RFP trip logic.

C. NONE of the RFPs will trip because the trip logic requires at least two (2) redundant trip signals.

D. All three (3) RFPs will trip because the Feedwater Level Control System will cause actual RPV level to rise above level 8.

LOC-23 NRC Exam Rev 3

K&A # 216000 K4.03 Importance Rating 3.4 QUESTION 30 RO Tier 2 Group 2 K&A Statement: Knowledge of NUCLEAR BOILER INSTRUMENTATION design feature(s) and/or interlocks which provide for the following:

Redundancy of sensors Justification:

A. Incorrect, no RFPs trip. Logic is 2 out of 3. Candidates may select this if they do not correctly recall the RFP trip logic arrangement.

B. Incorrect, logic is not arranged this way. Candidates may select this if they do not correctly recall the RFP trip logic arrangement.

C. Correct, the logic requires 2 out of 3 r'edundant level 8 signals to actuate the trip. This trip is applied to all 3 feed pumps when met.

D. Incorrect, the ICS/DCS FWLC system contains sufficient redundant RPV level Signals to be impervious to one level input failing, so actual RPV level will not falsely respond to the failure. Candidates may select this if they do not correctly recall that ICS/DCS will not respond to the level transmitter failure KIA Match Justification:

This question matches the stated KIA since candidates must recall how the RFP trip logic and ICS/DCS utilizes redundant sensors to monitor and control RPV water level.

References:

TM-OP-080 rev 9 Reference Required none Learning Objective: 10561.d Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 6/2110 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 31 Unit 2 is in mode 5 with the following conditions present:

• Reactor Mode Switch is in REFUEL

• Unit 2 Refueling platform Reactor Select Switch selected to NORM

• Unit 2 Refueling platform is positioned over the unit 2 reactor

• Refuel switch #1 is activated

• Fuel grapple is UNLOADED'

• Monorail hoist is UNLOADED

• Frame mounted hoist is UNLOADED

• Control Rod 10-23 is at position 48 Given these conditions, which one of the following changes would prevent reverse refueling platform motion?

A. The Frame mounted hoist is loaded.

B. Refuel Switch #2 is activated.

C. A second control rod is withdrawn beyond position 00.

D. Refuel switch #1 is DE-activated.

LOC-23 NRC Exam Rev 3

K&A# 234000 K5.01 Importance Rating 2.9 QUESTION 31 RO Tier 2 Group 2 K&A Statement: Knowledge of the operational implications of the following concepts as they apply to FUEL HANDLING EQUIPMENT:

Crane/hoist operation Justification:

A. Correct, reverse movement is blocked if selected to NORM and over NORM; a control rod is withdrawn; refuel switch #1 is activated; and EITHER fuel grapple loaded >550Ibs, OR frame hoist >500lbs; OR monorail hoist >500Ibs.

Raising the frame hoist load above this limit completes the REVERSE movement block circuit.

B. Incorrect, refuel switch #2 is not in this circuit and conditions to enable it to prevent reverse motion are not present. Candidates may select this if they incorrectly believe that refuel switch #2 provides input to the circuit.

C. Incorrect, an additional rod withdrawn will not affect the circuit and would not be permitted by RMCS. Candidates may select this if they incorrectly believe that a second control rod may be withdrawn in this condition.

D. Incorrect, de-activation of refuel switch #1 would indicate the bridge is no longer above the normal reactor and would permit reverse motion. Candidates may select this if they do not correctly recall the purpose and function of refuel switch #1.

KIA Match Justification:

This question matches the stated KIA since candidates must correctly recall details of fuel handling crane and hoist equipment operation.

References:

TM-OP-081 B rev 3 Reference Required none Learning Objective: 10787.C Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)5 Comments: Created by: T. North, 6/2110 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 32 Unit 1 is shutdown in mode 4 when a seismic event occurs.

As a result of the earthquake, the weld connecting the Unit 1 Main Condenser Hotwell Reject line to the "A" Condensate Storage Tank severs, and the reject line becomes completely detached from the CST tank wall.

The "A" Control Rod Drive pump is in service when the seismic event occurs.

This line break (1) affect the running CRD pump because (2)

A. (1) WILL NOT; (2) pump suction is re-aligned to the CRDH system return line when in mode 4 B. (1) WILL NOT; (2) pump cooling is re-aligned to the TBCCW system when in mode 4 C. (1) WILL; (2) net positive suction head for the pump will be lost D. (1) WILL; (2) cooling to the pump gearbox and bearings will be lost LOC-23 NRC Exam Rev 3

K&A # 201001 K6.01 Importance Rating 2.8 QUESTION 32 RO Tier 2 Group 2 K&A Statement: Knowledge of the effect that a loss or malfunction of the following will have on the CONTROL ROD DRIVE HYDRAULIC System:

Condensate system Justification:

A. Incorrect, pump suction cannot be re-ali~Jned, no other suction source is available.

Candidates may select this if they incorfi9ctly believe that the CRD suction path can be re-aligned.

B. Incorrect, pump cooling cannot be re-ali~lned, and is not provided by the condensate reject line. Candidates may select this if they incorrectly believe that the CRD suction path can be re-aligned.

C. Correct, the CRD pump suction is connected to the condensate reject line to provide NPSH. With this line disconnected from the CST, the pump no longer has NPSH.

D. Incorrect, pump cooling is normally provided by TBCCW and will not be affected by the loss of the CST reject connection. Candidates may select this if they incorrectly believe that the CST water provides cooling to the CRD pumps.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the physical interrelationship between condensate and CRD and predict the effect of a failure of the condensate connection.

References:

TM-OP-055 rev 5, M-146 rev 39 Reference Required none Learning Objective: 2419.b Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 6/3/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 33 Both SSES Units are operating at full power with Control Room Emergency Outside Air Supply System (CREOASS) in the following lineup:

• CREOASS Fan "A" (OV101A) is selected to AUTO-STANDBY

• CREOASS Fan "B" (OV1 01 B) is selected to AUTO-LEAD The mode switch for the "A" Outside Air Radiation Monitor, RISHH-D12 OK618A, is then placed to the TRIP TEST position.

Which one of the following is the correct response of the CREOASS system?

A. Fan "A" will automatically start, fan "B" will remain off.

B. Fan "B" will automatically start, fan "A" will remain off.

C. BOTH "A" AND "B" Fan auto start function on HI-HI outside air radiation is disabled.

D. ONLY the Fan "A" auto start function on HI-HI outside air radiation is disabled.

LOC-23 NRC Exam Rev 3

K&A # 290003 A 1.05 Importance Rating 3.2 QUESTION 33 RO Tier 2 Group 2 K&A Statement: Ability to predict and/or monitor changes in parameters associated with operating the CONTROL ROOM HVAC controls including: Radiation monitoring (control room)

Justification:

A. Correct, placing the rad monitor to TRIP TEST will auto start its associated divisional fan. The A monitor starts the A fan. Fan A will auto start regardless of the position of the auto-standby switch.

B. Incorrect, despite that the B fan is selected to lead, only the Brad monitor will start the B fan. Candidates may select this if they incorrectly believe that placing the A rad monitor to trip test will start the fan selected to lead.

C. Incorrect, the auto start feature for the fans is not disabled in trip test. Candidates may select this if they incorrectly believe that the trip test position will disable the fan auto start feature.

D. Incorrect, the auto start feature for the fans is not disabled in trip test. Candidates may select this if they incorrectly believe that the trip test position will disable the fan auto start feature.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the functional relationship between rad monitors and control room ventilation components and predict the effect of operating the rad monitor controls.

References:

TM-OP-030 rev 4, OP-030-002 rev 26 Reference Required none E-197-sheet 1 rev 16, -sheet 3 rev 22 Learning Objective: 1965.b Question source:  !\lew Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)5 Comments: Created by: T. North, 6/3/10 Reviewed by: T. Ebert, L Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 34 Unit 1 is operating in mode 1 at 70% power. Power ascension is in progress following refueling outage 17.

Control rod 38-31 is being withdrawn from position 12 to position 16 when a valid Rod Block Monitor (RBM) upscale trip halts rod motion.

Annunciator AR-103-C04, "RBM UPSCALE OR INOP ROD BLOCK" is illuminated.

In this condition, control rod motion stopped when RBM reached (1)

AND, in accordance with alarm response procedure AR-103-001-C04, operators must (2) in order to continue rod withdrawal.

A. (1) 117%

(2) select an edge rod to clear the rod block, then re-select rod 38-31 ONLY B. (1) 117%

(2) verify thermal limits will not be exceeded; de-select then re-select rod 38-31 C. (1) 109.2%

(2) select an edge rod to clear the rod block, then re-select rod 38-31 ONLY D. (1) 109.2%

(2) verify thermal limits will not be exceeded; de-select then re-select rod 38-31 LOC-23 NRC Exam Rev 3

K&A # 215002 A2.01 Importance Rating 3.3 QUESTION 34 RO Tier 2 Group 2 K&A Statement: Ability to (a) predict the impacts of the following on the ROD BLOCK MONITOR SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Withdrawal of control rod in high power region of core: BWR 3,4,5 Justification:

A. Incorrect, it is inappropriate in this condition to simply clear the rod block by selecting an edge rod. This action may work to temporarily clear the block, however, since the rod block was due to a valid local overpower condition, this action alone is insufficient to permit continued rod motion. Verification of thermal limit margin is required to ensure fuel damage will not occur due to the local power conditions.

Candidates may select this since the procedure permits this action when the rod block is spurious.

B. Correct, per the cycle 17 COLR, the intermediate setpoint applies when reactor power is >61 %. The alarm response procedure requires that the crew verify with RE that there is sufficient margin to thermal limits, then reselect the rod to reset the rod block (the rod must first be de-seleced, although the procedure does not state this directly).

C. Incorrect, the intermediate rod block setpoint of 117% is in effect when power is

>61 % and < 81 %. Candidates may select this if the do not correctly recall the RBM setpoint, and they assume they can continue rod motion without thermal limit verification (See A above).

D. Incorrect, the intermediate rod block setpoint of 117% is in effect when power is

>61 % and < 81 %. Candidates may select this if they do not correctly recall the intermediate RBM setpoint.

KIA Match Justification:

This question matches the stated KIA since candidates are required to recall the conditions necessary to provide a valid rod block when withdrawing a control rod in a central (high powered) region of the core. Additionally, they must recognize the appropriate action to be taken in accordance with the approved procedure for the stated condition of a valid rod block.

References:

COLRfT'RM section 3.2 rev 11; TM- Reference Required none OP-078K rev 4; AR-1 03-001-C04 rev 38.

Learning Objective: 15806.i, 15811.e Question source: Modified INPO bank #25961 Question History: Pilgrim 2003 NRC Exam Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)5 Comments: Modified by: T. North, 8/31/10 Reviewed by: T. Ebert, l. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 35 Unit 1 is in mode 2 with a reactor startup in progress.

The PCOM selects control rod 10-31 and momentarily depresses the W/DRAW ROD pushbutton to withdraw the rod one notch.

In order to determine if the COMPLETE rod withdraw sequence has taken place as designed, what indications should be observed, AND in what sequence?

A. (1) rod w/draw light momentarily lit (2) withdraw flow of 2-3 gpm (3) rod settle light lit, then off B. (1) rod insert light lit, then off (2) insert flow of 2-3 gpm (3) rod w/draw light lit, then off (4) withdraw flow of 4-5gpm (5) rod settle light lit, then off C. (1) rod w/draw light lit, then off (2) withdraw flow of 2-3 gpm (3) rod insert light momentarily lit (4) rod settle light lit, then off D. (1) rod insert light momentarily lit (2) rod w/draw light lit, then off (3) withdraw flow of 2-3 gpm (4) rod settle light lit, then off LOC-23 NRC Exam Rev 3

K&A # 201002 A3.02 Importance Rating 2.8 QUESTION 35 RO Tier 2 Group 2 K&A Statement: Ability to monitor automatic operations of the REACTOR MANUAL CONTROL SYSTEM including: Rod movement sequence lights Justification:

A. Incorrect, the insert light must be observed to momentarily light. Candidates may select this if they do not correctly recall the light sequence following rod withdrawal.

B. Incorrect, the insert light is lit only momentarily, and insert flow will not be observed.

Withdraw flow is normally 2-3 gpm, not 4-5. Candidates may select this if they do not correctly recall the light sequence following rod withdrawal and that withdraw flow is 2-3 gpm.

C. Incorrect, the insert light will momentarily light before the wid light. Candidates may select this if they do not correctly recall the light sequence following rod withdrawal.

D. Correct, OP-156-001 defines this as the correct indications for a one notch rod withdrawal sequence.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the correct sequence of rod movement lights that must be observed following control rod withdrawal.

References:

OP-156-001 rev 16 Reference Required none Learning Objective: 2469.d Question source: SSES OPS_INITIAL_LlCENSE Bank #TMOP056A12470 001 Question History: Bank Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)7 Comments: Created by:

Reviewed by: T. Ebert, L Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 36 Unit 2 has experienced a reactor scram following a trip of BOTH Reactor Recirc Pumps.

• Feedwater Pump A is in service for reactor water level control

• Reactor Water Cleanup (RWCU) is in letdown mode

• BOTH Reactor Recirc loops are NOT isolated Operators have been directed to monitor Reactor Pressure Vessel (RPV) water temperature to assess the status of RPV thermal stratification in accordance with ON-200-101, "Scram, Scram Imminent".

In addition to RPV saturation pressure and temperature, which one of the following temperature indications should be used for this purpose?

A. Reactor Recirc loop temperatures.

B. RWCU bottom head drain temperature.

C. Reactor Vessel wall aild flange temperatures.

D. Feedwater system RPV inlet temperature.

LOC-23 NRC Exam Rev 3

K&A # 204000 A4.09 Importance Rating 2.9 QUESTION 36 RO Tier 2 Group 2 K&A Statement: Ability to manually operate and/or monitor in the control room:

Reactor water temperature Justification:

A. Incorrect, ON-200-101 directs the use of bottom head drain temp. Recirc loop temperatures will not provide accurate RPV water level temps with recirc pumps off.

Candidates may select if they incorrectly believe that loop temperature indicators can provide accurate indication with the loops not isolated.

B. Correct. ON-200-101 directs that RWCU bottom head drain temperature be used to determine the status of thermal stratification since it provides the most appropriate and accurate temperature in the bottom head region necessary to assess the degree of RPV thermal stratification.

C. Incorrect, although these thermocouples may provide useful information regarding the surface temperature of the vessel itself, these temperatures are not relied upon in ON-200-101 to provide the status of thermal stratification. Candidates may select this if they incorrectly assume these indicators will provide information about stratified temperature layers on the vessel wall and flange.

D. Incorrect, feedwater RPV inlet temperature will not provide temperature indication for conditions internal to the RPV that are required to assess stratification conditions.

Candidates may select this if they believe this indicator will provide accurate temperature for the vessel downcomer and inlet plenum with the FW system in service.

KIA Match Justification:

This question matches the stated KIA since candidates are required to recall that RWCU system provides the most accurate and useful RPV water temperature indication needed to monitor for thermal stratification.

References:

ON-200-101 rev 19 Reference Required none Learning Objective: 1700.a Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)7 Comments: Created by: T. North, 8/3"1110 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 37 Which one of the following statements describes the purpose of Speed Limiter

1. 2 in the Reactor Recirc Flow Control system?

A. Limits recirc pump speed to 48% to lower reactor power to within the capacity of the main condenser in the event of a Circ water pump trip.

o B. Limits recirc pump speed ~?/% to lower reactor power to within the capacity of the main condenser in the event of a eirc water pump trip.

C. Limits recirc pump speed to 48% to ensure sufficient NPSH to jet pumps in the event of a low RPV water level condition.

D. limits recirc pump speed t~ #0 to ensure sufficient NPSH to jet pumps in the event of a low RPV water level condition.

LOC*23 NRC Exam Rev 3

K&A # 202002 2.1.28 Importance Rating 4.1 QUESTION 37 RO Tier 2 Group 2 K&A Statement: Recirc Flow Control:

Conduct of Operations: Knowledge of the purpose and function of major system components and controls.

Justification:

A. Correct, this is the setpoint and basis for speed limiter #2 per TM-OP-o64E.

B. Incorrect, this setpoint is recirc pump minimum speed. Limiter #2 will only run the recirc pump to 48%. Candidates may select this if they do not correctly recall the setpoint for limiter 2.

C. Incorrect, this is the basis for limiter #1. Candidates may select this if they incorrectly recall the limiter #2 basis.

D. Correct, this is the minimum speed setpoint and the limiter #1 basis. Candidates may select this if they incorrectly apply the speed limiter #1 basis and minimum speed setpoint to limiter #2.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the purpose and function of recirc pump speed limiters.

References:

TM-OP-064E rev 0 Reference Required none Learning Objective: 16021.d Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)7 Comments: Created by: T. North, 7/1/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 38 Which one of the following describes the correct sequence used when Reactor Building Zone 1 ventilation system is started in accordance with OP-134-002, "Reactor Building HVAC Zones 1 and 3", AND WHY this sequence is used?

A.

• The filtered exhaust fan is started first

• Once a negative pressure has been drawn the operator starts the supply and exhaust fans

• This sequence minimizes the pressure transient on the building.

B.

• The filtered exhaust fan is started first.

• Once a negative pressure has been drawn the operator starts the supply and exhaust fans.

• This sequence prevents an immediate trip of the supply fan on low flow.

C.

• The control switches for the supply and exhaust fans are first placed in start

• The filtered exhaust fan is then started, causing all three fans to start simultaneously.

• This sequence minimizes the pressure transient on the building.

D.

• The control switches for the supply and exhaust fans are first placed in start

• The filtered exhaust fan is then started, causing all three fans to start simultaneously.

• This sequence prevents an immediate trip of the supply fan on low flow.

LOC-23 NRC Exam Rev 3

K&A # 290001 K4.02 Importance Rating 3.4 QUESTION 38 RO Tier 2 Group 2 K&A Statement: Knowledge of SECONDARY CONTAINMENT design feature(s) and/or interlocks which provide for the following: Protection against over pressurization Justification:

A. Incorrect, the switches are interlocked to ensure they are operated simultaneously.

Candidates may select this if they do not recall that the fans are interlocked.

B. Incorrect, the switches are interlocked to ensure they are operated simultaneously to limit the pressure transient. Candidates may select this if they do not recall that the fans are interlocked, and do not correctly recall the basis for the interlock.

C. Correct. Each of these fans is interlocked with the other so the switches are operated in this manner to start all off one switch change. This interlock prevents building overpressure by ensuring an exhaust fan is running when a supply fan is running.

D. Incorrect, the purpose of the interlock is to limit the building pressure transient.

Candidates may select this if they do not correctly recall the basis for the interlock.

KIA Match Justification:

This question matches the stated KIA since canclidates are required to recall knowledge of the fan starting interlock for RB zone 1 ventilation, and its basis.

References:

OP-134-002 rev 47 Reference Required none Learning Objective: 1274.n Question source: SSES OPS_INITIAL_LlCENSE Bank #TMOP034/1277 001 Question History: Bank Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41(b)7 Comments: Created by: Bank Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 39 Unit 2 is operating in mode 4, with RHR pump C in shutdown cooling.

RHR pump C then TRIPS.

• ALL other RHR pumps are UNAVAILABLE

• BOTH Reactor Recirc Pumps are UNAVAILABLE

• RPV Level is +75", steady Per ON-249-001, "Loss Of RHR Shutdown Cooling Mode", RPV level is currently (1) to promote natural circulation because it is (2)

A. (1) TOO LOW (2) BELOW the top of the steam separators B. (1) HIGH ENOUGH (2) ABOVE the top of the steam separators C. (1) TOO LOW (2) BELOW the bottom of the steam dryer skirt D. (1) HIGH ENOUGH (2) ABOVE the bottom of the steam dryer skirt LOC-23 NRC Exam Rev 3

K&A# 295001 AK1.01 Importance Rating 3.5 QUESTION 39 RO Tier 1 Group 1 K&A Statement: Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Natural circulation Justification:

A. Incorrect, RPV level at or above 45" promotes natural circulation since level will be above the top of the steam seperators. Therefore, level is currently sufficient since it is below the required level. Candidates may select this if they incorrectly recall the internal location of RPV components and/or the RPV level required to allow natural circ to take place.

B. Correct, level must be above +45" and the top of the steam separators. With level at +75" the top of the steam separators will be covered.

C. Incorrect, level is currently above the bottom of the dryer skirt (approximately 0"),

however this is not the level required to promote natural circ which is significantly higher. Candidates may select this if they incorrectly recall the internal location of RPV components and the RPV level required to allow natural circ to take place.

D. Incorrect, although level is above the bottom of the steam dryer skirt, this is not high enough to allow circulation to take place, since level must be above the steam separators. Candidates may select this if they incorrectly recall the RPV level required to allow natural circ to take place.

KIA Match Justification:

This question matches the stated KIA since candidates must determine the operation implications of current RPV level with respect to the promotion of natural circulation from a procedural and physical standpoint following a complete loss of all forced circulation.

References:

ON-149-001 rev 23 Reference Required none Learning Objective: 15310.0 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41(b)3 Comments: Created by: T. North, 9/7/10 Reviewed by: T. Ebert, l. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 40 Unit 1 has been operating at full power for the past year. A half scram is in on RPS Channel B when the "A" High Drywell Pressure Scram instrument, PSH C72-N002A, fails UPSCALE.

Which one of the following describes plant conditions approximately 10 seconds later?

Reactor thermal power generation will be approximately (1) of full power, and this heat will be removed via the (2)

A. (1) 7%

(2) Safety Relief Valves B. (1) 1%

(2) Bypass Valves C. (1) 100%

(2) Main Turbine Control Valves D. (1) 7%

(2) Bypass Valves LOC-23 NRC Exam Rev 3

K&A# 295006 AK1.01 Importance Rating 3.7 QUESTION 40 RO Tier 1 Group 1 K&A Statement: Knowledge of the opE~rational implications of the following concepts as they apply to SCRAM: Decay heat generation and removal Justification:

A. Incorrect, There is no condition that would result in an MSIV isolation, and BPVs are available. Candidates may select this if they incorrectly believe that bypass valves will not be available.

B. Incorrect, decay heat will be approximatl~ly 7%. Candidates may select this if they cannot correctly recall the amount of decay heat following a full scram from 100%

power.

C. Incorrect, the instrument failure will cause a full scram. Candidates may select this if they do not correctly determine that a full scram will occur.

D. Correct, The instrument failure will result in a full scram combined with the half scram in trip system B; and deca'y heat following a scram from full power will be approximately 7% after 10 seconds. Since there is no condition that would cause an MSIV isolation, turbine bypass valves will be available to control pressure and remove decay heat.

KIA Match Justification:

This question matches that stated KIA since candidates must first determine that a full scram will occur, and then determine the operational impact of the generation of decay heat following this scram.

References:

TM-OP-058 rev 9 Reference Required none Learning Objective:

Question source: Modified II\lPO Bank #25978 Question History: 2003 Pilgrim NRC exam Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)8 Comments: Modified by: T. North, 10**3-10 Reviewed by: T. Ebert, L.. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 41 Complete the following statement regarding a degraded voltage condition on 480 VAC safety busses:

Protection against a LOW voltage condition in the Reactor Protection power distribution system is provided by the (1) , because the low voltage condition may cause (2)

A. (1) RPS motor generator set voltage regulator; (2) scram pilot solenoids to chatter and potentially lose the ability to actuate when required.

B. (1) RPS motor generator set voltage regulator; (2) instrument setpoints to drift in a NON-conservative direction affecting their scram safety functions.

C. (1) Electrical Protection Assembly Breaker; (2) scram pilot solenoids to chatter and potentially lose the ability to actuate when required.

D. (1) Electrical Protection Assembly Breaker; (2) instrument setpoints to drift in a NON-conservative direction affecting their scram safety functions.

LOC-23 NRC Exam Rev 3

K&A# 295003 AK1.03 Importance Rating 2.9 QUESTION 41 RO Tier 1 Group 1 K&A Statement: Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF A.C. POWER: Under voltage/degraded voltage effects on electrical loads Justification:

A. Incorrect, while the RPS MG set provides voltage regulation, it is not relied upon for protection against an undervoltage condition. It also provides no ability to control the voltage in the alternate supply. Candidates may select this if they do not recall which component protects against the UV condition.

B. Incorrect, while the RPS MG set provides voltage regulation, it is not relied upon for protection against an undervoltage condition. It also provides no ability to control the voltage in the alternate supply. Instrument setpoint drift is not the basis for the UV trip. Candidates may select this if they do not recall which component protects against the UV condition, or the resultant effect.

C. Correct, the low voltage trip of the EPA breakers in both the normal and alternate RPS supply is designed to provide the undervoltage protection for this condition. Per TSB 3.3.8.2 "In the event of a low voltage condition for an extended period of time, the scram solenoids can chatter and potentially lose their pneumatic control capability, resulting in a loss of primary scram function."

D. Incorrect, instrument setpoint drift is not the basis for the UV trip. Candidates may select this if they do not correctly recall the operational implication of the UV condition on RPS busses.

KIA Match Justification:

This question matches the stated KIA since candidates must understand the operational implications of a degraded voltage condition on 480 VAC busses and the component that prevents this.

References:

TM-OP-058 rev 9, TSB 3.3.8.2 Reference Required none Learning Objective: 10071.b, 15970 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)8 Comments: Created by: T. NOl1h, 5/26/10 Reviewed by: T. Ebert, L Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUES"nON 42 Unit 1 is operating at full power when a grid disturbance causes the Main Generator output breaker to trip.

• The "Turbine Control Valve Fast Closure Scram" instruments fail to initiate a Reactor Protection System (RPS) scram

• RPV pressure peaks at 1106 psig until bypass valves regain pressure control

• APRM power peaks at 102%

• NO operator action has been taken Which one of the following describes the status of control rods following this event?

A. All control rods inserted due to Alternate Rod Insertion actuation on high RPV pressure.

B. All control rods inserted due to automatic RPS actuation on high RPV pressure.

C. Control rods will NOT insert until operators arm & depress the Manual Scram pushbuttons on 1C651 .

D. Control rods will NOT insert until opE~rators arm & depress the Alternate Rod Insertion pushbuttons on 1C601.

LOC-23 NRC Exam Rev 3

K&A # 295025 EK2.01 Importance Rating 4.1 QUESTION 42 RO Tier 1 Group 1 K&A Statement: Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: RPS Justification:

A. Incorrect, the ARI automatic setpoint on high pressure has not been exceeded, and since the high pressure scram and BPVs function properly, ARI auto initiation will not occur. Candidates may select this if they incorrectly believe the ARI setpoint has been exceeded and do not recall that the RPS high pressure scram will actuate.

B. Correct, RPV pressure has exceeded the scram setpoint for high RPV pressure, resulting in all rods in.

C. Incorrect, auto scram will occur on high pressure. Candidates may select this if they do not recall that the pressure transient will cause an automatic scram.

D. Incorrect, auto scram will occur on high pressure. Candidates may select this if they do not recall that the pressure transient will cause an automatic scram.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the relationship between high reactor pressure and RPS scram actuation setpoints.

References:

TM-OP-058 Reference Required none Learning Objective: 2486.a Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 {b)7 Comments: Created by: T. North, 617/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 43 A LOCA has occurred at Unit 1.

• The PCOP is operating Residual Heat Removal (RHR) pumps in LPCI mode to maintain RPV water level above -129".

• NO other sources of injection to the RPV are available.

• A suppression pool (SP) leak then occurs resulting in RHR pump operation BELOW the RHR pump vortex limit.

The PCOP can expect to be directed to ...

A. stop injection and secure ALL RHR pumps until SP level can be restored.

B. secure RHR pumps ONLY if SP temperature rises causing a further reduction in net positive suction head.

C. continue RPV injection with RHR pumps, but limit RHR flow to LESS THAN 7000 gallons per minute.

D. continue RPV injection with RHR pumps with NO restrictions.

LOC-23 NRC Exam Rev 3

K&A# 295030 EK2.04 Importance Rating 3.7 QUESTION 43 RO Tier 1 Group 1 K&A Statement: Knowledge of the interrelations between LOW SUPPRESSION POOL WATER LEVEL and the following: RHR/LPCI Justification:

A. Incorrect, injection should not be stoppE~d since RHR is required to ensure adequate core cooling. Candidates may select this if they incorrectly determine that exceeding vortex limits will preclude use of RHR for this purpose.

B. Incorrect, a rise in SP temp may likely reduce RHR pump NPSH, however it is not evaluated in EOPs for the purpose of restricting RHR pump operation. Further, RHR pumps are required to assure adequate core cooling and may continue to inject without restriction per EOP guidance.

C. Incorrect, unlike core spray, the RHR vortex limit is a straight line, below which any flowrate is not permitted, UNLESS RHR pumps are required to maintain RPV level.

Candidates may select this if they incorrectly believe flow restrictions apply in this condition.

D. Correct, since RHR pumps are required for adequate core cooling this takes precedence over exceeding vortex limits, and RHR injection may continue without restriction.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the relationship between suppression pool level and operation of RHR pumps and correctly apply the procedural vortex limit.

References:

EO-1 00-1 03 rev 9, EO-000-103 rev 7 Reference Required none Learning Objective: 14616 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 11-14-10 Reviewed by: E. Brice, A. Avery, 11-23-10 LOC*23 NRC Exam Rev 3

QUESTION 44 Unit 1 is operating at full power, when a steam line rupture occurrs inside the drywell.

• Drywell temperature is 275°F, up fast

• Drywell pressure is 4.5 psig, up slow

• The crew has determined that drywell sprays are required In this situation, which condition must the crew observe, AND why?

A. Limit initial drywell spray flow to between 1000 and 2800 gpm to prevent excessive evaporative cooling that could damage primary containment internal components and structures.

B. Do not start drywell sprays until suppression chamber pressure exceeds 13 psig to prevent excessive evaporative cooling that could damage primary containment internal components and structures.

C. Limit initial drywell spray flow to between 1000 and 2800 gpm to prevent the cyclic condensation of steam at the downcomer openings of the drywell vents.

D. Do not start drywell sprays until suppression chamber pressure exceeds 13 psig to prevent the cyclic condensation of steam at the downcomer openings of the drywell vents.

LOC-23 NRC Exam Rev 3

K&A # 295028 EK2.02 Importance Rating 3.2 QUESTION 44 RO Tier 1 Group 1 K&A Statement: Knowledge of the interrelations between HIGH ORYWELL TEMPERATURE and the following: Components internal to the drywell Justification:

A. Correct, the spray flow limit is applicable in this instance and prevents an excessive evaporative cooling pressure drop that could challenge the drywell to suppression chamber dp limits, and damage primary containment components or structure.

B. Incorrect, While this statement is true, this limit is not applicable in this scenario, since the decision to spray is based solely on OW temperature trend. Candidates may select this if they incorrectly determine that OW sprays may not be initiated until Supp Chmbr pressure exceeds 13 psig.

C. Incorrect, the cyclic steam condensation at the downcomer (chugging) is the reason drywell sprays are not started until 13 psig when spraying from the PC/P leg of the PC control EOP, and is N/A for this condition. Candidates may select this if they incorrectly determine that OW sprays may not be initiated until Supp Chmbr pressure exceeds 13 psig.

O. Incorrect, drywell sprays should be started to limit the OW temperature excursion.

The chugging phenomenon does not apply in this situation. Candidates may select this if they do not correctly recall the basis for the OW sprays flow limit.

KIA Match Justification:

This question matches the stated KIA since candidates must interrelate the potential for internal component damage with required actions taken to mitigate a high drywell temperature condition.

References:

EPG rev 2, SSES-PSTG rev 8 Reference Required none Learning Objective: 14613 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 5/213/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 45 Which one of the following describes the reason for the 'Drywell Pressure-High' function for the Reactor Protection System Instrumentation?

A. Decrease the probability of exceeding primary containment design limits following a complete loss of drywell cooling.

B. Prevent the loss of equipment inside the drywell needed for accident mitigation following a complete loss of drywell cooling.

C. Prevent the loss of equipment inside the drywell needed for accident mitigation following a break in the Reactor Coolant Pressure Boundary.

D. Decrease the probability of fuel damage following a break in the Reactor Coolant Pressure Boundary.

LOC-23 NRC Exam Rev 3

K&A# 295024 EK3.06 Importance Rating 4.0 QUESTION 45 RO Tier 1 Group 1 K&A Statement: Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: Reactor SCRAM Justification:

A. Incorrect, see below. Candidates may select this if they do not correctly recall the basis for the high drywell pressure scram.

B. Incorrect, see below. Candidates may select this if they do not correctly recall the basis for the high drywell pressure scram.

C. Incorrect, see below. Candidates may select this if they do not correctly recall the basis for the high drywell pressure scram.

D. Correct, per TSB 3.3.1.1; "High pressure in the drywell could indicate a break in the RCPB. A reactor scram is initiated to minimize the possibility of fuel damage and to reduce the amount of energy being added to the coolant and the drywell. The Drywell Pressure-High Function is assumed in the analysis of the recirculation line break (Ref. 6). The reactor scram reduces the amount of energy required to be absorbed and, along with the actions of Emergency Core Cooling Systems (ECCS), ensures that the fuel peak cladding temperature remains below the limits of 10 CFR 50.46."

KIA Match Justification:

This question matches the stated KIA since candidates must recall knowledge of the reason for generating an automatic reactor scram following a high drywell pressure transient.

References:

TSB 3.3.1.1 rev 4 Reference Required none Learning Objective: 15970 Question source: INPO bank #21805, Perry 2001 NRC exam Question History: Bank Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41(b)7 Comments: Created/Modified by:

Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 46 Complete the following statement:

The operator actions specified in ON-013-001, "Response to Fire", are intended to ...

A. define the specific firefighting tactics and requirements in the event of a fire in areas outside the control room ONLY.

B. define the specific firefighting tactics and requirements in the event of a fire in areas inside AND outside the control room.

C. establish safe plant shutdown and cooldown capabilities in the event of a fire in areas outside the control room O~ILY.

D. establish safe plant shutdown and cooldown capabilities in the event of a fire in areas inside AND outside the control room.

LOC-23 NRC Exam Rev 3

K&A# SOOOOO AK3.04 Importance Rating 2.8 QUESTION 46 RO Tier 1 Group 1 K&A Statement: Knowledge of the reasons for the following responses as they apply to PLANT FI RE ON SITE: Actions contained in the abnormal procedure tor plant fire on site Justification:

A. Incorrect, specific firefighting tactics and requirements are defined in pre-fire plans established for each fire zone. Candidates may select this if they do not correctly recall the reasons for actions specified in ON-01-001.

B. Incorrect, specific firefighting tactics and requirements are defined in pre-fire plans established for each fire zone. Candidates may select this if they do not correctly recall the reasons for actions specified in ON-01-001.

C. Correct, ON~013-001 establishes a method and identifies available eqUipment to meet appendix R safe shutdown paths for fires in zones outside the control room only. Fires inside the control room are not addressed in this procedure.

D. Incorrect, Fires inside the control room are not addressed in this procedure. The control room abandonment procedure provides the necessary steps for fires in the control room. Candidates may select this if they do not correctly recall the reasons for actions specified in ON-01-001.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the basiS for actions contained in the fire on site abnormal procedure, ON-013-001, "Response To Fire".

References:

ON-013-001 rev 28 Reference Required none Learning Objective: 15310.b Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41(b)10 Comments: Created by: T. North, Sn/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 47 Complete the following statement regarding Extraction Steam Non-Return Valves:

Following a Main Turbine trip, Extraction Steam Non-Return Valves (Bleeder Trip Valves) are (1) in order to (2L--.

A. (1) OPEN (2) divert steam flow from the turbine to prevent a turbine overspeed.

B. (1) SHUT (2) prevent a turbine overspeed by stopping reverse steam flow from the feedwater heaters.

C. (1) OPEN (2) divert steam flow from the turbine to allow a faster turbine coastdown.

D. (1) SHUT (2) stop the flow of extraction steam to feedwater heaters to raise core inlet subcooling.

LOC-23 NRC Exam Rev 3

K&A # 295005 AK3.05 Importance Rating 2.5 QUESTION 47 RO Tier 1 Group 1 K&A Statement: Knowledge of the reasons for the following responses as they apply to MAIN TURBINE GENERATOR TRIP: Extraction steam/moisture separator isolations Justification:

A. Incorrect, NRVs are shut on a turbine trip because the residual steam will reverse and continue to turn the turbine. Candidates may select this if they incorrectly believe that NRVs open following turbinE~ trip.

B. Correct, NRVs shut to prevent turbine overspeed caused by the steam from feedwater heaters reversing and continuing to drive the main turbine with no electrical load.

C. Incorrect, NRVs are shut on a turbine trip. Coastdown is related to friction created by turbine and generator bearings and main condenser vacuum, and will be relatively unaffected by extraction steam flow. Candidates may select this if they incorrectly believe that NRVs open following turbine trip.

D. Incorrect, although RPV inlet temperature will indeed drop, the reason for stopping extraction steam to FW heaters is not related to core inlet subcooling since the reactor will be shutdown in this condition. Candidates nay select this if they incorrectly recall the basis for isolating I\IRVs after turbine trip.

KIA Match Justification:

This question matches the stated KIA since canclidates must recall the basis for isolating extraction steam following a main turbine trip.

References:

TM-OP-093 rev 10 Reference Required none Learning Objective: 1614.f Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41(b)7 Comments: TM-OP- Created by: T. North, 5/~~6/1 0 093 refers to NRVs Reviewed by: T. Ebert, L. Casperson 11-12-10 as "Extraction Steam Non-Return Valves";

OP-193-001 refers to NRVs as "Bleeder Trip Valves" therefore BOTH terms are included in the question stem.

LOC-23 NRC Exam Rev 3

QUESTION 48 Unit 1 has experienced a LOCA resulting in entry into EO-1 00-1 02, RPV Control, and EO-1 00-1 03, PC Control.

The following conditions exist:

• Suppression Pool level is 20 feet and steady

• The PCOP is controlling RPV pressure using Safety Relief Valve manual operation, resulting in rising suppression pool temperature The Unit Supervisor directs the PCOP to report suppression pool temperature.

Which one of the following describes the instrumentation available to the PCOP to accurately determine Suppression Pool temperature?

t~ wtvt- iLTf)s OAfl; A. SPOTMOS Division~ 1-aRd. 2 ave-FaQ~ temperature.

B. SPOTMOS Division 1 average temperature ONLY.

c. PICSY Division 1 and 2 Bulk Temperature A (MAT 37).

D. SPOTMOS Division 1 lower RTDs ONLY.

LOC-23 NRC Exam Rev 3

K&A # 295026 EA 1.03 Importance Rating 3.9 QUESTION 48 RO Tier 1 Group 1 K&A Statement: Ability to operate and/or monitor the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE:

Temperature monitoring Justification:

A. Incorrect, these sensors will not be submerged and will not provide an accurate value. Candidates may select this if they do not correctly recall which temperature indicators are accurate with low SP leVE!1.

B. Incorrect, these sensors will not be submerged and will not provide an accurate value. Candidates may select this if they do not correctly recall which temperature indicators are accurate with low SP level.

C. Incorrect, this computer data utilizes sensors tAM are not submerged, therefore will C)'f not be accurate. Candidates may select this if they do not correctly recall which temperature indicators are accurate with low SP level.

D. Correct, the sixteen RTDs are located near the water surface at a level of 20.5 feet above Suppression Pool bottom. Normal Suppression Pool operating level is 23 feet - minimum level is 22 feet. For Division I only, four additional RTDs provide input to SPOTMOS. "rhese RTDs (TE 15751, TE-15756, TE-15761, and TE-15764) are located deep in the pool, three feet above Suppression Pocil bottom.

KIA Match Justification:

This question matches the stated KIA since candidates must determine which SP level monitoring indication is appropriate with high SP temp and low SP level.

References:

TM-OP-059Z rev 5 Reference Required None Learning Objective: 330.a Question source: SSES OPS_INITIAL_LlCENSE Bank #TMOP059Z1330 003 Question History: Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created/Modified by:

Reviewed by: T. Ebert, L Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 49 Unit 1 has experienced an accident resulting in fuel damage and a radioactive release via the Turbine Building Stack.

• All unit 1 SPING field units are UNAVAILABLE due to a loss of Instrument AC panel1Y219 Which one of the following describes the USE~ of the Post Accident Vent Stack Sampling System (PAVSSS) in this condition:

The PAVSSS ...

A. CANNOT be utilized to monitor the release, since the PAVSSS field units can ONLY be used to monitor the Reactor Building Stack.

B. CAN be utilized to monitor the release and can provide BOTH noble gas AND particulate concentrations.

C. CAN be utilized to monitor the release but can ONLY provide noble gas concentration.

D. CANNOT be utilized to monitor the release since the PAVSSS field units can ONLY be used to monitor the Standby Gas Treatment Exhaust.

LOC*23 NRC Exam Rev 3

K&A # 295038 EA 1.05 Importance Rating 3.0 QUESnON49 RO Tier 1 Group 1 K&A Statement: Knowledge of the reasons for the following responses as they apply to HIGH OFF-SITE RELEASE RATE: Post accident sample system (PASS): Plant-Specific Justification:

A. Incorrect, the PAVSSS can only monitor the TB stack and SGTS vent. Candidates may select this if they do not correctly recall the purpose and function of PAVSSS monitors.

B. Incorrect, the PAVSSS cannot monitor particulate concentrations. Candidates may select this if they do not correctly recall the purpose and function of PAVSSS monitors.

C. Correct, the PAVSSS is designed to be a backup to the SPING under accident conditions and its ability to monitor the stack release will not be affected by the power loss. It can only provide noble gas concentration.

D. Incorrect, the PAVSSS stack monitoring components will not be affected by the loss of 1Y219. Candidates may select this if they do not correctly recall the power supply to PAVSSS monitors.

KIA Match Justification:

This question matches the stated KIA since candidates must recall facts regarding use and response of PAVSSS monitors during a high off site release event.

References:

TM-OP-079Z rev 4 Reference Required none Learning Objective: 10396.b Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 9/1 H/1 0 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 50 Unit 1 is in a refueling outage. Unit 2 is at 100% power. A refueling accident has occurred on 818' level resulting in the following annunciators alarming on Unit 1:

• REFUEL FLOOR WALL EXH HI RADIATION (AR-112-D1)

• REFUEL FLOOR WALL EXH HI-HI RADIATION (AR-101-AS)

Which one of the following describes the Standby Gas Treatment (SGTS) system response (if any) to the event?

A. BOTH SGTS trains start and align to Zone III.

B. BOTH SGTS trains start and align to Zone I AND Zone III.

c. ONLY ONE SGTS train starts and aligns to Zone III.

D. NO SGTS trains start and ventilation remains in a normal lineup.

LOC-23 NRC Exam Rev 3

K&A # 295023 AA 1.07 Importance Rating 3.6 QUESTION 50 RO Tier 1 Group 1 K&A Statement: Ability to operate and/or monitor the following as they apply to REFUELING ACCIDENTS: Standby gas treatment Justification:

A. Correct, the high high refuel floor wall rad monitor setpoint has been exceeded and results in the auto start of BOTH SGTS trains and alignment to zone 3 only.

B. Incorrect, SGTS aligns to zone 3. Candidates may select this if they do not correctly recall that SGTS aligns only to zone 3 upon this initiation signal.

C. Incorrect, both SGTS trains start. Candidates may select this if they do not recall that both SGTS trains will respond.

D. Incorrect, SGTS auto start will occur. Candidates may select this if they do not correctly recall that SGTS will start and align to Zone 3 following only the refuel floor wall hi-hi rad.

KIA Match Justification:

This question matches the stated KIA since candidates must correctly recall the SGTS system response in order to properly monitor SGTS automatic action following the annunciators.

References:

TM-OP-070 rev 5 Reference Required none Learning Objective: 1991 Question source: SSES NRC Bank #6 Question History: Bank Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)7 Comments: Created/Modified by: Bank Reviewed by: T. Ebert, L. Casperson 11-12-10 LQC-23 NRC Exam Rev 3

QUESTION 51 Unit 1 was operating at full power when a LOCA concurrent with a loss of all high pressure feed occurred resulting in the following conditions:

• RPV level is -208", steady

• NO RHR pumps are available

• Core Spray loop A is injecting to the RPV at 6500 gpm

• Core Spray loop B is NOT injecting to the RPV

• RPV pressure is 75 psig, down slow

• 6 ADS SRVs are open

• RPV injection has been maximized using ALL available systems Which one of the following statements describes the status of Adequate Core Cooling, and why?

A. Adequate Core Cooling is NOT assured; RPV Water level is below Top of Active Fuel.

B. Adequate Core Cooling IS assured; Core Spray loop A flow is above 6350 gpm C. Adequate Core Cooling is NOT assured; Core Spray loop B flow is below 6350 gpm.

D. Adequate Core Cooling IS assured; RPV pressure is above the Minimum Steam Cooling Pressure.

LOC-23 NRC Exam Rev 3

K&A # 295031 EA2.04 Importance Rating 4.6 QUESTION 51 RO Tier 1 Group 1 K&A Statement: Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: Adequate core cooling Justification:

A. Incorrect, adequate core cooling IS assured even though RPV level is below TAF since CS A is injecting >6350 gpm. The candidate may choose this if they incorrectly believe that ACC cannot be assured when level is below TAF.

B. Correct, with RPV level below -161" but above -210", adequate core cooling can only be assured by spray cooling**.at least one CS loop above design flow of 6350 gpm C. Incorrect, ACC IS assured. Core spray flow needs to be ~ 6350 gpm in at least one loop alone. As long as CS A is >6350 gpm, CS B flow is not required to assure ACC.

The candidate may choose this if they incorrectly believe that both CS loops must be injecting to assure ACC.

D. Incorrect, The MSCP value only applies during ATWS conditions when RPV level is undetermined. In this case MSCP is irrelevant and does not factor into ACC determination. The candidate may choose this if they incorrectly believe that MSCP is a relevant determinant to ACC.

KIA Match Justification:

This question matches the stated KIA since candidates are required to evaulate current plant conditions with RPV water level below TAF to determine if adequate core cooling can be assured.

References:

EO-000-102 rev 8 Reference Required none Learning Objective: 14591 Question source: New Question History: New Cognitive level: Memory!Fundamental knowledge:

Comprehension!Analysis: X 10CFR55 41 (b)10 Comments: Created by: T. North, 8!~-I1!10 Reviewed by: T. Ebert, l.. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUES1"ION 52 Unit 1 is at 100% power when a leak develops in the Reactor Building Chilled Water (RBCW) discharge piping.

If RBCW flow continues to degrade due to the leak with NO operator action, determine which one of the following describes the Reactor Building Closed Cooling Water (RBCCW) system response:

A. RBCCW will provide cooling flow to the Reactor Recirc Pump motor winding coolers IMMEDIATELY AFTER RBCW drops below 1 psid.

B. RBCCW will provide cooling flow to the Drywell coolers AFTER RBCW drops below 1 psid for 13 seconds.

C. RBCCW flow to the Reactor Water Cleanup NON-regenerative Heat Exchanger (NRHX) ISOLATES IMMEDIATELY AFTER RBCW drops below 1 psid.

D. RBCCW flow to the Reactor Building Sump Cooler ISOLATES AFTER RBCW drops below 1 psid for 13 seconds.

LOC*23 NRC Exam Rev 3

K&A# 295018 AA2.04 Importance Rating 2.9 QUESTION 52 RO Tier 1 Group 1 K&A Statement: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: System flow Justification:

A. Incorrect, RBCCW assumes cooling flow to the RRP winding coolers occurs after a 13 second time delay. Candidates may select this if the do not correctly recall the details of the low flow transfer signal.

B. Correct, the flow degradation will result in transfer of cooling for RBCW drywell loads to RBCCW 13 seconds after RBCW flow drops below 1psid. RBCCW flow to the NRHX will also be isolated by this signal.

C. Incorrect, RBCCW cooling to the NRHX will isolate upon the low RBCW system flow signal after a 13 second time delay. Candidates may select this if the do not correctly recall the time delay portion of system r'9sponse.

D. Incorrect, RBCCW will continue to provide flow to the sump cooler regardless of the status of RBCW. Candidates may select this if the do not correctly recall that sump cooling flow is unaffected by the system response.

KIA Match Justification:

This question matches the stated KIA since candidates must determine how a reduction in RBCW system flow will be impact the RBCW and RBCCW systems.

References:

TM-OP-014 rev 3, TM-OP-300 rev 3, Reference Required none ON-134-001 rev 26 Learning Objective: 1694.d Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)10 Comments: Created by: T. North, 10/18/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 53 Unit 1 was operating at full power when a control room fire occurred requiring evacuation of the control room in accordance with ON-1 00-009, "Control Room Evacuation" .

• All remote shutdown transfer switches have been repositioned

• The control room has been successfully evacuated

• NO systems, structures or components have been damaged as a result of the fire

• RPV pressure is 1040 psig, steady

• RPV level is+35", steady Direction has been given to depressurize thl~ RPV per ON-1 00-009.

Given ON-100-009 attachments A and B, which one of the following is the CORRECT action operators should take to accomplish this?

A. ADS SRVs G, J, K, L, M, or N should be operated from the Upper Relay Room to reduce RPV pressure to LESS THAN 100 psig as soon as possible to allow RHR to be placed in Shutdown Cooling.

B. ADS SRVs G, J, K, L, M, or N should be operated from the Upper Relay Room to reduce RPV pressure to NO LESS THAN 400 psig over the next hour.

C. SRVs A, B or C should be operated from the Remote Shutdown Panel to reduce RPV pressure to LESS THAN 100 psig as soon as possible to allow RHR to be placed in Shutdown Cooling.

D. SRVs A, B or C should be operated from the Remote Shutdown Panel to reduce RPV pressure to NO LESS THAN 400 psig over the next hour.

LOC*23 NRC Exam Rev 3

K&A # 295016 AA2.03 Importance Rating 4.3 QUESTION 53 RO Tier 1 Group 1 K&A Statement: Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Reactor pressure Justification:

A. Incorrect, ADS valves should only be operated if SRVs A, Band C are unavailable.

No condition is present allowing the CDR to be exceeded. Candidates may select this if they incorrectly determine that CDR may be exceeded and SRVs A,B and C may be used.

S. Incorrect, ADS valves should only be operated if SRVs A, Band C are unavailable.

Candidates may select this if they incorrectly determine that SRVs A,S and C may be used.

C. Incorrect, No condition is present allowing the CDR to be exceeded. Candidates may select this if they incorrectly determine that CDR may be exceeded.

D. Correct, ON-100-009 directs operators to cooldown using SRVs A, B, and C.

The cooldown should be conducted less than 1OO"Flhr since there is no condition present requiring RD or anticipation of RD. Limiting the pressure drop to 400 psig over the next hour keeps CDR below TS limits.

KIA Match Justification:

This question matches the stated KIA since candidates must determine the change in pressure necessary to depressurize in accordance with the procedure provided and within required limits.

References:

ON-100-009 rev 21 Reference Required: ON-100 009 Att A&B Learning Objective: 15306, 15307 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41(b)10 Comments: Created by: T. l\Iorth, 6/24/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 54 Unit 1 is in mode 5, with RHR pump "A" in Shutdown Cooling (SDC).

A leak develops on the "A" RHR pump suction piping resulting in rising water level in the "A" RHR pump room, and lowering reactor cavity water level.

• AR-109-H08, "RHR LOOP A PUMP ROOM FLOODED", alarm is illuminated

• Operators manually close the SDC inboard and outboard isolation valves PRIOR to reaching the SDC auto isolation signal on low RPV water level

• RPV water level stabilizes at +15"

• The leakage stops when the SDC isolation valves are shut Which one of the following identifies the procedures that REQUIRE IMMEDIATE ENTRY?

A. ON-149-001, "Loss of RHR Shutdown Cooling Mode" ONLY B. ON-149-001, "Loss of RHR Shutdown Cooling Mode"; AND EO-1 00-104, "Secondary ContainmEmt Control", ON LY C. ON-149-001, "Loss of RHR Shutdown Cooling Mode"; AND EO-100-102, "RPV Control", ONLY D. ON-149-001, "Loss of RHR Shutdown Cooling Mode"; AND EO-100-102, "RPV Control"; AND EO-1 00-1 04, "Secondary Containment Control" LOC-23 NRC Exam Rev 3

K&A # 295021 2.4.1 Importance Rating 4.6 QUESTION 54 RO Tier 1 Group 1 K&A Statement: Emergency Procedures / Plan: Knowledge of EOP entry conditions and immediate action steps.

Justification:

A. Incorrect, SC Control EOP requires entry. Candidates may select this if they do not recognize the SC Control EOP entry.

B. Correct, SC Control EOP requires entry on hi level in the RHR pump room. No entry condition exists for RPV Control.

C. Incorrect, 1\10 entry for RPV control, SC Control entry exists on hi RHR pump room level. Candidates may select this if they do not recognize the SC Control EOP entry and incorrectly determine RPV Control must be entered.

D. Incorrect, 1\10 entry for RPV control. Candidates may select this if they incorrectly determine RPV Control must be entered.

KIA Match Justification:

This question matches the stated KIA since candidates must evaluate current conditions and determine which EOP entry conditions have been met. Boiling Water Reactor EOPs do not have "immediate action steps" since aI/ actions must be directed by the Control Room Supervisor after carefully evaluating plant status. Therefore, the immediate action steps portion of the KIA was not addressed.

References:

EO-000-102 rev 8, 104; ON-149-001 Reference Required None rev 23 Learning Objective: 14585 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41(b)10 Comments: At Created by: T. North, 6/9/10 SSES, EOPs do not Reviewed by: T. Ebert, L. Casperson 11-12-10 have immediate action steps.

LOC-23 NRC Exam Rev 3

QUESTION 55 Unit 1 is operating at full power when a failure of the Div 1 +24 VDC Battery Charger 10673 occurs.

As a result of this failure, Div 1 24 VDC Panel 10672 wilL ..

A. remain ENERGIZED from the (-)24 VDC battery charger 10674.

B. remain ENERGIZED from battery bank 10670.

C. be DE-ENERGIZED until operators can place the (-)24 VDC battery charger 10674 in service.

D. be DE-ENERGIZED until operators can reposition the Alternate Power Transfer Switch HS-111501.

LOC-23 NRC Exam Rev 3

K&A # 295004 AK2.01 Importance Rating 3.1 QUESTION 55 RO Tier 1 Group 1 K&A Statement: Knowledge of the interrelations between PARTIAL OR COMPLETE LOSS OF D.C. POWER and the following: Battery Charger Justification:

A. Correct, the 24 vdc bus is normally powered by a positive and negative battery charger. The battery is divided into a positive and negative bank. Per "rM-OP 075, upon failure of the positive charger, the negative charger will assume the full load.

B. Incorrect, the (-) charger will remain in service powering the DC panel. Candidates may select this if they do not correctly H~call 24 vdc distribution details.

C. Incorrect, the panel will remain energized via the + battery and (-) charger.

Candidates may select this if they do not correctly recall 24 vdc distribution details.

D. Incorrect, the panel will remain energized. The alternate power transfer switch only switches ac power to both chargers simultaneously and will not need to be repositioned. Candidates may select this if they do not correctly recall 24 vdc distribution details.

KIA Match Justification:

This question matches the stated KIA since candidates must recall knowledge of battery charger and DC bus response following failure of a battery charger.

References:

TM-OP-075 rev 2 Reference Required none Learning Objective: 10102 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 11122110 Reviewed by:

LOC-23 NRC Exam Rev 3

QUES"nON 56 Unit 1 is in Mode 4 with preparations for a reactor startup in progress.

A seismic event occurs resulting in a complete loss of instrument air and a small LOCA inside the drywell.

Given the following plant conditions:

• RPV level is +5", down slow

• Condensate system was in long path recire prior to the seismic event Which one of the following statements is CORRECT regarding the RPV level control strategy?

Condensate pumps ...

A. CANNOT be used to feed the RPV because the Condensate Pump Discharge Valves have failed shut.

B. CAN be used to feed the RPV, BUT FLOW CANNOT BE THROTTLED since The Low Load Flow Control and the Low Load Bypass Valves have FAILED SHUT. .

C. CAN be used to feed the RPV since the air loss will NOT affect the low load flow control and bypass valves.

D. CANNOT be used to feed the RPV since the Short Path Recirc AND Feed Pump Min Flow Recire valves have FAILED OPEN.

LOC-23 NRC Exam Rev 3

K&A # 295019 2.4.9 Importance Rating 3.8 QUESTION 56 RO Tier 1 Group 1 K&A Statement: Loss of Instrument Air:

Emergency Procedures / Plan: Knowledge of low power /

shutdown implications in accident (e.g., loss of coolant accident or loss of residual heat removal) mitigation strategies.

Justification:

A. Incorrect, condensate pump discharge valves are motor operated and will not be affected by the air loss. Candidates may select this if they do not correctly recall this fact.

B. Incorrect, condensate will be unavailable due to the short path recirc and disch vents open. Long path recirc valves also fail open. Candidates may select this if they do not correctly recall these facts. '

C. Incorrect, the low load FCV and bypass valve fail shut on loss of air. Candidates may select this if they do not correctly recall this fact.

D. Correct, the loss of IA will result in the short path and RFP min flow recirc valves failing open preventing the use of condensate pumps in this condition.

KIA Match Justification:

This question matches the stated KIA since candidates must determine how the loss of air impacts EOP mitigating strategies during shutdown conditions.

References:

ON-118-001 rev 23, TM-OP-044 rev 8, Reference Required none

-045 rev 13 Learning Objective: 1823.a Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)1 0 Comments: Created by: T. North, 6/21/10 Reviewed by: T. Ebert, L. Casperson 11-12-1 0 LOC-23 NRC Exam Rev 3

QUESTION 57 Which one of the following statements is CORRECT regarding 4KV ESS busses when abandoning the Control Room per ON-1 00-009, "Control Room Evacuation"?

A. ALL ESS busses should be transferred to their respective Emergency Diesel Generators by operating the diesels and switchgear locally.

B. ALL ESS busses should remain energized via their normal source unless offsite power is lost to the startup transformers.

C. ONLY the Division 1 ESS busses should be transferred to their respective Emergency Diesel Generators by operating the diesels and circuit breakers locally.

D. ONLY the Division 2 ESS busses should be transferred to their respective Emergency Diesel Generators by operating the diesels and circuit breakers locally.

LOC-23 NRC Exam Rev 3

K&A # 295016 AA1.04 Importance Rating QUESTION 57 RO Tier 1 Group 1 K&A Statement: Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT: A.C. electrical distribution Justification:

A. Incorrect, ESS busses are allowed to remain powered from the S/U XFMRS.

Candidates may select this if they incorrectly believe ESS busses should be transferred.

B. Correct, the ON does not direct any action to transfer power to diesels unless offsite power is lost.

C. Incorrect, ESS power remains from S/U XFMRS. Candidates may select this if they incorrectly believe ESS busses should be transferred.

D. Incorrect, ESS power remains from Stu XFMRS. Candidates may select this if they incorrectly believe ESS busses should be transferred.

KIA Match Justification:

This question matches the stated KIA since candidates must recall how the AC distribution system is operated when abandoning the control room.

References:

ON-100-009 rev 21 Reference Required none Learning Objective: 15310.d Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41(b)7 Comments: Created by: T. North, 7/1110 Reviewed by: T. Ebert, L Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 58 Unit 1 is operating at full power when disturbances in the electrical grid result in entry into ON-198-001, "Unit 1 Main Generator MVAR Control For Auto Voltage Regulator Operation When Synched To Grid".

Given the data obtained by the crew and recorded on ON-198-001, page 3, and attachment "A", "Main Generator Reactive Capability Curve":

(the data obtained was taken 5 minutes ago, and is unchanged)

Determine the CORRECT action for this condition AND why:

A.

• RAISE GENERATOR MW OUTPUT;

• This will cause a corresponding reduction in reactive load to within the capability curve.

B.

• RAISE GENERATOR EXCITATION;

• This will raise the allowable reactive load by increasing the lagging power factor.

C.

• REDUCE GENERATOR EXCITATION;

• This will reduce the reactive load to within the capability curve.

D.

• REDUCE HYDROGEN GAS PRESSURE;

• This will shift the capability curves to allow additional reactive load.

LOC*23 NRC Exam Rev 3

K&A# 700000 AK3.02 Importance Rating 3.6 QUESTION 58 RO Tier 1 Group 1 K&A Statement: Knowledge of the reasons for the following responses as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Actions contained in abnormal operating procedure for voltag '9 and grid disturbances.

Justification:

A. Incorrect, raising generator MW would not be permissible in this case since it would require exceeding reactor license MWth limit. Additionally, this action would not have the desired effect, and would bring conditions further outside the capability curve.

Candidates may select this if they do not fully understand the actions required to maintain generator parameters within the capability curve.

B. Incorrect, raising generator eXCitation would raise generator voltage and thereby raise the reactive load further outside the capability curve. Candidates may select this if they do not fully understand the actions required to maintain generator parameters within the capability curve.

C. Correct, since the combination of reactive load and MW output exceed the capacity curve, reducing generator excitation will reduce generator voltage and the corresponding reactive load will be reduced. This action is directed by the ON to bring reactive load to within the capability curve.

D. Incorrect, reducing generator gas pressure would make reactive load limits more restrictive since generator cooling capability would be reduced, and is not a desired action in this case. Candidates may selElct this if they do not fully understand the actions required to maintain generator parameters within the capability curve.

KIA Match Justification:

This question matches the stated KIA since candidates must recall knowledge of the reasons why actions required by the off normal procedure are taken following grid disturbances.

References:

ON-198-001 rev 11, TM-OP-098 rev 5 Reference Required ON-198 001 page 3 of 16 with data, and attachment A

Learning Objective: 15306, 15307, 15318 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41(b)4,5,7,10 Comments: Created by: T. North, 7/3/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 59 OSCAR has been dispatched as a result of a refueling accident on the refuel floor (818 1). The Standby Gas Treatment System (SGTS) automatically initiates.

The following conditions exist:

• Zone 1 and III differential pressure is -0.31 inches WG.

• SGTS SPING Noble Gas is 1.0E06 micro curies per minute.

• OSCAR whole body dose readings are 0.05 mRem/hour.

A siding panel fails on the Refuel Floor. Zone III differential pressure now indicates 0 inches WG.

(1) How do SPING readings relate to the offsite release rate; AND (2) How will OSCAR whole body dose readings respond to the panel failure?

A. (1) SBGT SPING Noble Gas IS representative of the Total Offsite Release.

(2) OSCAR whole body dose readings will NOT change.

B. (1) SBGT SPING Noble Gas IS representative of the Total Offsite Release.

(2) OSCAR whole body dose readings will increase.

C. (1) SBGT SPING Noble Gas is NOT representative of the Total Offsite Release.

(2) OSCAR whole body dose readings will NOT change.

D. (1) SBGT SPING Noble Gas is NOT representative of the Total Offsite Release.

(2) OSCAR whole body dose readings will increase.

LOC-23 NRC Exam Rev 3

K&A # 295035 EK1.01 Importance Rating 3.9 QUESTION 59 RO Tier 1 Group 2 K&A Statement: Knowledge of the operational implications of the following concepts as they apply to SECONDARY CONTAINMENT HIGH DIFFERENTIAL PRESSURE: Secondary Containment Integrity Justification:

A. Incorrect, the SC integrity failure results in a release bypassing SGTS and causing dose rates to rise. Candidates may select this if they do not consider the impact of the SC siding failure.

B. Incorrect since the SC integrity failure causes SGTS to be bypassed resulting in SGTS sping readings not indicative of total release rates. Candidates may select this if they do not consider the impact of the SC siding failure.

C. Incorrect, OSCAR readings will rise as release rate increases. Release rate increases through the siding failure. Candidates may select this if they do not consider the impact of the SC siding failure.

D. Correct, with SC integrity no longer intact, and dP high, radioactive material released due to the refueling accident will bypass SGTS and cause site dose rates as indicated by OSCAR to rise.

KIA Match Justification:

This question matches the stated KIA since candidates must determine the operational impact due to the rise in offsite dose following a secondary containment failure due to a rise in secondary containment pressure.

References:

TM-OP-034 rev 7 Reference Required none Learning Objective: 1266.a, b Question source: SSES NRC Exam Bank :11:650 Question History: SSES 2005 NRC Exam Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41(b)8 Comments: Created/Modified by:

Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 60 Unit 1 is operating in mode 1 with reactor power at 75%. A primary leak into secondary containment is in progress.

Radiation levels in the reactor building are rising, with the following Area Radiation Monitor indications:

4

• ARM channel 50, "CRD NORTH" indicates 2 x 10 MRlHR

• ARM channel 52. "RWCU RECIRC PP ACC" indicates 3 x 104 MRlHR ~

• All other ARM channels are reading between 100 and 500 IVIRlHR Refer to EO-100-004, "Secondary Containment Control", table 9 below and determine which one of the following describes the impact of these conditions on plant operation:

TABLES REACTOR BUilDING RADIATION MAX NORMAl. MAX SAFE ReRAn ARM NUM6 R RAmATlON RAOfA'OON (RlHR)

FIElO Flao EO 1(14 R8AR1':A LO HfGH ARM CHANNa a(Fr} RANGE RANGE OESCRIPTIGi (MRlI-R) (RIHR) 8HI Ja. loll... CASK ITOft AMiA HIIALARM ro" 10 I .... loll... SPENT fUB. CRiT MOH HI" loll'" RERII!t. fU)Oft HOR1'lI 42+ loll'" RERII!t. ft.OOft weST 47' loll... SPENT FUB. CRiT MOH loll... 4S RERII!t. Ft.OOft AMiA

,/ ,

7.

II<> \'~

~/ RWCU RECIRC PP ACe 4

til" $4 fUlB. 1'00.. PP AREA HI~ 10 lG ft+ HIA me IIU) SAMII'I..e ST

~ ---.......

l~

6" CROHOR'TH 119- 4 Ii' 51 CROlOOTK HIIAIARU 10 10 The crew is required to insert a reactor scram AND ...

A. stabilize RPV pressure below 1087 psig.

B. perform Rapid Depressurization using ADS SRVs.

c. anticipate Rapid Depressurization using turbine bypass valves.

D. cooldown the RPV within normal cooldown rates.

LOC*23 NRC Exam Rev 3

K&A# 295033 EK2.01 Importance Rating 3.8 QUESTION 60 RO Tier 1 Group 2 K&A Statement: Knowledge of the operational implications of the following concepts as they apply to HIGH SECONDARY CONTAINMENT AREA RADIATION L.EVELS : Area radiation monitoring system Justification:

A. Incorrect, a rapid depressurization usin~) ADS is also required. Candidates may select this if they do not recall that 2 are!as above max safe values requires RD with a primary leak in to SC.

B. Correct, scram AND RD with ADS valves are both required since 2 ARMs from 2 separate areas are reading above max safe values.

C. Incorrect, depressurization may not be performed with BPVs in this case since conditions requiring RD with ADS valves are met. The candidate may select this if they do not recall that use of BPVs is not permitted.

D. Incorrect, RD exceeding cooldown rates is required. The candidate may select this if they confuse actions required when rad levels are caused by conditions other than a primary leak into SC.

KIA Match Justification:

This question matches the stated KIA since candidates are required to determine the impact on plant operation of area rad monitor readings above max safe values.

References:

EO-100-104, rev 7, EO-000-104, rev 6 Reference Required none Learning Objective: 14586.m, 14594 Question source: INPO Bank #22275 Question History: Nine Mile Point U2 2002 NRC exam Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 8/::l1/10 Reviewed by: T. Ebert, L.. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 61 Under ATWS conditions, operators are directed by EO-1 00-113, "Level/Power Control", sheet 2, Control Rod Insertion, to bypass the Rod Worth Minimizer (RWM).

Which one of the following is the reason for this action?

The RWM will prevent inward rod motion due to an abnormal rod pattern if reactor power ...

A. remains ABOVE the Low Power Setpoint (LPSP) ONLY.

B. remains ABOVE BOTH the Low Power Setpoint (LPSP) AND Low Power Alarm Point (LPAP).

C. drops BELOW the Low Power Setpoint (LPSP) ONLY.

D. drops BELOW the Low Power Alarm Point (LPAP) ONLY.

LOC-23 NRC Exam Rev 3

K&A # 295015 AK3.01 Importance Rating 3.4 QUESTION 61 RO Tier 1 Group 2 K&A Statement: Knowledge of the reasons for the following responses as they apply to INCOMPLETE SCRAM: Bypassing rod insertion blocks Justification:

A. Incorrect, the RWM will not enforce the sequence unless power is BELOW the LPSP. Candidates may select this if they misunderstand the relationship between LPSP, LPAP and sequence enforcement.

B. Incorrect, the RWM will not enforce the sequence with power above both the LPAP and LPSP. Candidates may select this if they misunderstand the relationship between LPSP, LPAP and sequence enforcement.

C. Correct, the RWM rod block features are enabled only when power is below the LPSP. These features may enforc:e a control rod insert block if rods are selected that are not in accordance with the loaded sequence. The RWM must be manually bypassed to ensure that rod blocks are not activated when rod insertion is attempted.

D. Incorrect, the RWM will not enforce the rod sequence until power is below the LPSP.

Power may have dropped below the LPAP, but the RWM will not enforce the sequence and stop rod motion. Candidates may select this if they misunderstand the relationship between LPSP, LPAP and sequence enforcement.

KIA Match Justification:

This question matches the stated KIA since candidates are required to recall the conditions that provide the basis for bypassing the rod worth minimizer and its related rod insertion blocks under incomplete scram conditions.

References:

EO-OOO-113 rev 8, TM-OP-031 D rev 4 Reference Required none Learning Objective: 14613 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)7 Comments: Created by: T. North, 8/31/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 62 Unit 1 is operating at 100% when a spurious Generator Load Reject results in anATWS.

The following conditions are present

• Reactor power is 69~steady

• MSIVs are OPEN Which one of the following statements describes the RPV pressure response to this event?

A. Bypass Valves will open fully, and RPV pressure will rise until Safety Relief Valves open to limit the pressure rise.

B. Bypass Valves ALONE will automatically throttle steam flow and maintain RPV pressure stable.

C. Bypass Valves AND Turbine Control Valves will automatically throttle steam flow and maintain RPV pressure stable.

D. Bypass Valves AND Turbine Control Valves will remain shut and RPV pressure will rise until Safety Relief Valves open to limit the pressure rise.

LOC-23 NRC Exam Rev 3

K&A# 295007 AA1.05 Importance Rating 3.7 QUES'nON 62 RO Tier 1 Group 2 K&A Statement: Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: Reactor/turbine pressure regulating system Justification:

A. Correct, current reactor power is beyond the total steam flow capacity of BPVs. EHC will cause BPVs to fully open, but RPV pressure will rise and require the use of SRVs to gain control of RPV pressure.

B. Incorrect, the steam flow demand will cause BPVs to fully open, and is beyond the capacity of BPVs to throttle and control pressure alone. Candidates may select this if they donot correctly recall the full capacity of turbine bypass valves.

C. Incorrect, TCVs will be shut as a result of the load reject scram. Candidates may select this if they incorrectly assume that turbine control valves will receive an open signal to assist in pressure control.

D. Incorrect, BPVs will be open. Candidates may select this if they incorrectly assume that bypass valves will be shut KIA Match Justification:

This question matches the stated KIA since candidates are required to evaluate plant conditions to determine the correct response of the reactor/turbine pressure regulating system. This enables proper monitoring of the system response following the rise in reactor pressure resulting from A TWS power exceeding bypass valve capability.

References:

TM-OP-093 rev 9, -093L rev 6 Reference Required none Learning Objective: 1641.0 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41(b)7 Comments: Created by: T. North, 6/21/10 Reviewed by: T. Ebert, L. Casperson 11-12-1 0 LOC*23 NRC Exam Rev 3

QUESTION 63 The Unit 1 Reactor Building Stack Monitor Panel - Rad Measurement 1C216B (29-818') alarm horn has actuated AND the Green High light is illuminated.

Which one of the following identifies a source of airborne radioactivity that would result in this panel alarming?

A. Unit 1 Turbine Building Ventilation Exhaust. I B. Standby Gas Treatment Exhaust.

C. Zone 2 Ventilation Exhaust.

D. Unit 1 Zone 3 Ventilation Exhaust.

LOC-23 NRC Exam Rev 3

K&A # 295034 EA2.02 Importance Rating 3.7 QUESTION 63 RO Tier 1 Group 2 K&A Statement: Ability to determine and/or interpret the following as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Cause of high radiation levels Justification:

A. incorrect - would alarm U1 TB stack. Candidates may select this if they are unfamiliar with the potential causes of local radiation alarms.

B. incorrect - would alarm SBGT stack alarm. Candidates may select this if they are unfamiliar with the potential causes of local radiation alarms.

C. incorrect - would alarm U2 stack alarm. Candidates may select this if they are unfamiliar with the potential causes of local radiation alarms.

D. Correct, Unit 1 zone 3 ventillation exhausts thru unit 1 reactor building stack.

KIA Match Justification:

This question matches the stated KIA since candidates must determine the cause of the secondary containment high radiation alarm.

References:

TM-OP-034 rev 7, TM-OP-079Z rev 4 Reference Required none Learning Objective: 1942.a Question source: INPO Bank # 23810 Question History: SSES 2002 NRC Exam Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)10 Comments: Created by: Bank Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 64 Unit 1 was operating at 35% power when a small steam leak inside the Drywell (DW) occurred. The following conditions exist:

• Reactor manually scrammed, all rods in

• DW temperature is 140°F, up slow

• DW pressure is 0.9 psig, up slow

• RPV level dropped to + 15", and is now +35", steady The Unit Supervisor directs that all available drywell cooling be placed in service.

To accomplish this, the PCO must:

A. Reset the DW cooling logic isolation ONLY.

B. Reset the DW cooling logic isolation then place additional DW coolers to START LOW.

C. Place additional DW coolers to STAF~T HIGH ONLY.

D. Place the DW cooler keylock test switch to TEST LOCA, then place additional DW coolers to START LOW.

LOC-23 NRC Exam Rev 3

K&A# 295012 2.1.23 Importance Rating 4.4 QUESTION 64 RO Tier 1 Group 2 K&A Statement: High Orywell Temperature:

Conduct of Operations: Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Justification:

A. Incorrect, a OW cooling logic isolation has not occurred. Candidates may select this if they incorrectly believe a OW cooling logic isolation has occurred.

B. Incorrect, see A above C. Correct, since there is no LOCA signal present, additional coolers can be started in high speed.

O. Incorrect, no LOCA signal is present, therefore repositioning the TEST LOCA switch is not required. Candidates may select tlhis if they incorrectly believe that a LOCA signal is present.

KIA Match Justification:

This question matches the stated KIA since candidates must possess the ability to perform required procedural actions during a high dryweU temperature transient.

References:

OP-160-001 rev 13, TM-OP-073 rev 2 Reference Required none Learning Objective: 10419.a Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41(b)10 Comments: Created by: T. North, 6/30/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 65 Unit 1 was operating at full power when a false High Drywell Pressure signal was received, resulting in a reactor scram and HPCI injection into the RPV.

One (1) minute after the scram, the followin~~ conditions are present:

• All rods at position 00

• Both Reactor Recirc Pumps are running at 30% speed

• RPV level is +56"

• RPV pressure is 820 psig

• HPCI and Reactor Feed Pumps have tripped

• Immediate operator actions have been performed Which one of the following can the crew expect to occur over the next 10 minutes with NO additional operator action?

A. RPV level will rise due to decay heat generation; RPV pressure will remain constant due to bypass valve operation.

B. RPV level AND pressure will BOTH drop due to bypass valve operation.

C. RPV level will rise and RPV pressure will drop due to continued CRD pump flow.

D. RPV level AND pressure will BOTH rise due to decay heat generation.

LOC-23 NRC Exam Rev 3

K&A # 295008 AA2.04 Importance Rating 3.1 QUESTION 65 RO Tier 1 Group 2 K&A Statement: Ability to determine and/or interpret the following as they apply to HIGH REACTOR WATER LEVEL: Heatup rate: Plant-Specific Justification:

A. Incorrect, bypass valves will not open until RPV pressure rises to approximately 934 psig. Candidates may select this if they are unfamiliar with post scram RPV water level, pressure and temperature interaction, and the causes.

B. Incorrect, bypass valves will not open until RPV pressure rises to approximately 934 psig. Candidates may select this if they are unfamiliar with post scram RPV water level, pressure and temperature interaction, and the causes.

C. Incorrect, the cold water injected by CRD will not be sufficient to overcome the decay heat immediately following the scram. Candidates may select this if they are unfamiliar with post scram RPV water level, pressure and temperature interaction, and the causes.

D. Correct, with no operator action, decay heat will cause both level and pressure to rise slowly KIA Match Justification:

This question matches the stated KIA since candidates must determine the effect of post scram decay heat on RPV level and pressure after level has been allowed to rise to level 8.

References:

ON-1 00-1 01 rev 24 Reference Required none Learning Objective: 15300 Question source: New New Question History: New New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)5 Comments: Created by: T. North, 6/7/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 66 It is currently 10:30 AM on DAY shift, and th*e PCOM requires a SHORT TERM (30 minutes) shift relief of the Operator At The Controls (OATC) duties in order to attend a very brief meeting with Operations department management. The PCOM will be turning the OATC duties over to a licensed operator from the work control center.

Which one of the following describes the PROCEDURALLY REQUIRED elements for this S~':'-Shift relief?

A. verba~nover, panel walkdown, reviewpf current plant conditions AND~omplete,turnover sheets.

B. Verbal turnover, panel walkdown AND review of current plant conditions ONLY.

C. Verbal turnover AND review of curre t D. Verbal turnover, panel walkdown AND comp ete ONLY. l\..

LOC-23 NRC Exam Rev 3

K&A # 2.1.3 Importance Rating 3.7 QUESTION 66 RO Tier 3 K&A Statement: Knowledge of shift or short-term relief turnover practices.

Justification:

A. Incorrect, turnover sheets are not required. Candidates may select this if they incorrectly believe that turnover sheets are required in this situation.

B. Correct, per OP-AD-002 7.4.S.b.1, all turnover elements are required except completion of turnover sheets.

C. Incorrect, panel walkdown is also requir,ed. Candidates may select this if they do not correctly recall turnover required elements.

D. Incorrect, review of current plant conditions is also required. Candidates may select this if they do not correctly recall turnoVl3r required elements.

KIA Match Justification:

This question matches the stated KIA since candidates must recall required elements of a proper short term mid-shift turnover per OP-AD-002.

References:

OP-AD-002 rev 34 Reference Required none Learning Objective: 4086 Question source: Modified INPO Bank #19050 Question History: Clinton 2000 NRC exam Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)1 0 Comments: Modified by: T. North, 10/26/10 Reviewed by: T. Ebert, L.. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 67 What alternate methods of determining valvl9 position may be utilized when inaccessibility prevents physical operation or observation of indication?

A. Obtaining their positions as noted on the most current Status Control Log.

B. Verifying system parameters (flOW, pressure, etc) are as expected for the current plant conditions.

C. Noting the inaccessible valves for verification on the next planned or unplanned entry.

D. Referring to the most recently completed checkoff list on the system.

LOC*23 NRC Exam Rev 3

K&A# 2.1.29 Importance Rating 4.1 QUESTION 67 RO Tier 3 K&A Statement: Knowledge of how to conduct system lineups, such as valves, breakers, switches, i tC.

Justification:

A. Incorrect, the valves are not necessarily under Status Control. Candidates may select this if they do not realize that not all valve positions are tracked under status control.

B. Correct per OP-OOo-002, Valves, section 6.9 C. Incorrect, this does not provide current valve position indication. Candidates may select this if they do not understand that this will not provide the current valve position as required by procedure D. Incorrect, this is not procedurally directed and is not a "positive" method of determining valve position as required by procedure. Candidates may select this if assume that this is a "positive" method of position verification.

KIA Match Justification:

This question matches the stated KIA since candidates are required to recall procedural requirements for the conduct of valve lineups.

References:

OP-000-002 rev 8 Reference Required none Learning Objective: 14829 Question source: SSES OPS_INITIAL_L1CENSE Bank # AD044/14829 Question History: Bank Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41(b)10 Comments: Created/Modified by:

Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 68 Which condition below constitutes a Technical Specification SAFETY LIMIT violation, AND what action must be taken?

A. THERMAL POWER at 11% with reactor pressure at 650 psig;

~ thermaJ power belo¥l the limit withi! J 24 liOtlTS'"O'NLY.

B. Reactor steam dome pressure at 1,350 psig: [) t/ /)1;;:

Reduce pressure below the limit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> ONLY. AN*} ~l"::t-A( v:~',  ;-L .

tv /v,t!£ "v1"1'A~ '30 ~y C. THERMAL POWER at 3995 MWth for 60 minutes.

Immediately reduce power to below the limit AND report the violation to the NRC within 30 days.

D. THERMAL POWER at 30% with core flow at 6 Mlbm/hr.

Reduce core thermal power below the limit AND insert all control rods within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

, ()C*23 NRC Exam Rev 3

K&A# 2.2.22 Importance Rating 4.0 QUESTION 68 ROTier3 K&A Statement: Knowledge of limiting conditions for operations and safety limits Justification:

A. Incorrect, TS SL has not been violated, thermal power must be::;; 23% when <785 #

or < 10 MLBlhr. Violations of TS SLs require actions to be taken within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Candidates may select this if they confuse 10 mlb/hr with 23% power as stated in TS, and cannot correctly recall the required actions.

B. Incorrect, stated pressure is above the SL, however, compliance with TS SL AND insertion of control rods must be completed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Candidates may select this if they cannot correctly recall required actions, C. Incorrect, exceeding MWth license limit is not a safety limit. Candidates may select this if they confuse license power limits and TS SLs.

D. Correct, thermal power is >23% while flow is <10 MLBlhr which is a violation of SL 2.1.1.1. Per TS 2.2, compliance with the TS SL and insertion of all control rods must be completed within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

KIA Match Justification:

This question is a match to the stated KIA since candidates must recall safety limit required values and evaluate current plant parameters to determine compliance.

References:

TS 2.0 rev 4 Reference Required none Learning Objective: 13427, 13429 Question source: SSES OPS_INITIAL_L1CENSE Bank# TMOP401/0000 004 Question History: Bank Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)5 Comments: Modified by: T. North, 5/27/10 Reviewed by: T. Eblart, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 69 Using Standby Liquid Control System (SBlC) Piping and Instrumentation Diagram M-148, determine the effect of the following on SBlC system operation:

• Unit 1 is operating at full power when SBlC pressure control valve PCV 14811C (grid location B-2) fails SHUT.

Unit 1 SBlC system is ...

A. NOT capable of performing its intended safety function because the loss of air sparge will allow sediment to clog SBlC pump suction lines.

B. capable of performing its intended safety function, but remote SBlC tank level indication will NOT be available.

C. NOT capable of performing its intended safety function because the remote SBlC tank level alarms will prevent pump operation.

D. capable of performing its intended safety function, but sparging air will NOT be available for tank chemical additions.

LOC*23 NRC Exam Rev 3

K&A# 2.2.15 Importance Rating 3.9 QUESTION 69 RO Tier 3 K&A Statement: Ability to determine the expected plant configuration using design and configuration control documentation, such as drawings, line ups, tag-outs, etc.

Justification:

A. Incorrect, SBLC is still capable of injecting boron to the RPV, the sparge air is not lost, and is not used to prevent suction line clogging. Candidates may select this if they mis-read the print and incorrectly determine that sparge air is lost.

B. Correct, the failure of the PCV isolates air to the remote tank level indicators.

These instruments will not affect the ability of SBlC to inject boron. local ultrasonic level indication will be available.

C. Incorrect, the level alarms are not inputs to the SBLC start circuit. Candidates may select this if they do not correctly recall that the loss of the level alarms will not impact sblc pump operation.

D. Incorrect, the sparging air is not affected by the PCV failure. Candidates may select this if they mis-read the print and incorrectly determine that sparge air is lost KIA Match Justification:

This question matches the stated KIA since candidates must evaluate and determine SBLC system status and configuration using a controlled station print.

References:

M-148 rev 39, TM-OP-053 rev 9 Reference Required none Learning Objective: 1217.h Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41(b)10 Comments: Created by: T. North, 5/27/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 70 Reactor power is 80% and being returned to 100% power following special testing and a control rod sequence exchange. The following alarms are received in the control room.

• AR-015-D04, "STACK MONITORING SYS OC630/0C677 HI-HI RADIATIOI\J"

• AR-015-E04, "STACK MONITORING SYS OC630/0C677 HI RADIATION" Further investigation reveals that Turbine Building Exhaust Radiation (Point #5) is the cause of the alarm and Offgas subtrain flow is 75% of the value it was before the power increase began. Offgas recombiner flow has increased as power has increased.

Given ON-070-001 , "Abnormal Gaseous Radiation Release/CAM Alarms", what actions are required for this situation?

A. Isolate the Primary Coolant Degasifier.

B. Start the Common Offgas Recombiner and shutdown the Unit 1 Offgas recombiner.

C. Shutdown Radwaste Ventilation.

D. Isolate the failed open pair of OFFGAS DELAY LINE DRAIN VLVS.

LOC-23 NRC Exam Rev 3

K&A# 2.3.11 Importance Rating 3.8 QUESTION 70 RO Tier 3 K&A Statement: Ability to control radiation releases.

Justification:

A. incorrect - this action would be appropriate ONLY after chemistry sampled the degasifiers and determined that they are the source of the high radiation ..

B. incorrect - with the drop in offgas flow, a candidate may believe the source of the problem to be with the recombiners, in which case shutting down the ineffective recombiner and starting the common recombiner would be appropriate C. incorrect - appropriate if Radwaste is believed to be the source, but this is not consistent with the indications given the drop in offgas flow.

D. Correct, Per ON-070 001 step 3.2 this is the probable cause given the drop in w

offgas flow KJA Match Justification:

This question matches the stated KJA since candidates determine the correct method to control the radiation release based on plant conditions and the procedure provided

References:

ON-070-001 rev 16 Reference Required none Learning Objective: 15308 Question source: SSES NRC Bank #364 Question History: SSES 2003 Cert exam Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)11 Comments: Created/Modified by: Bank Reviewed by: A. Avery, E. Brice 11-23-10 LOC*23 NRC Exam Rev 3

QUESTION 71 Unit 1 is in MODE 5 with the drywell open for maintenance.

In order to operate the TIP system in accordance with OP-178-001, "TIP System", under these conditions the Health Physics department must restrict access to the:

TIP room AND ...

A. Drywell AND GIG mezzanine ONLY.

B. Drywell AND north HCU area ONLY.

G. GIG mezzanine ONLY.

D. GIG mezzanine, AND north HCU arE~a ONLY.

LOC-23 NRC Exam Rev 3

K&A# 2.3.12 Importance Rating 3.2 QUESTION 71 RO Tier 3 K&A Statement: Knowledge of Radiological Safety Principles pertaining to licensed operator duties, such as containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Justification:

A. Correct per OP-178-001 B. Incorrect, north HCU area access is not restricted. Candidates may select this if they are unfamiliar with areas affected by elE~vated rad levels in the TIP room.

C. Incorrect, drywell access is also restrictl3d. Candidates may select this if they are unfamiliar with areas affected by elevatE~d rad levels in the TIP room.

D. Incorrect, north HCU area access is not restricted, and DW access is controlled.

Candidates may select this if they are unfamiliar with areas affected by elevated rad levels in the TIP room.

KIA Match Justification:

This question matches the stated KIA since candidates must recall radiological principles associated with high radiation levels in the TIP room (locked high rad area).

References:

OP-178-001 rev 19 Reference Required none Learning Objective: 10152 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41(b)12 Comments: Modified by: T. North, 6/22110 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 72 Which one of the following identifies conditions of thermal hydraulic instability in the core following a Reactor Recirc pump trip from SO% RTP per ON-17S-002, "Core Flux Oscillations"?

A. APRM "A" Simulated Thermal Power rod block occurs and immediately clears.

B. APRM peak to peak oscillations are approximately 7% and rising.

C. LPRM upscale alarms occur and clear 10 seconds later.

D. Recirc loop flow peak to peak oscillations are 3% and steady.

LOC*23 NRC Exam Rev 3

K&A# 2.4.21 Importance Rating 4.0 QUESTION 72 RO Tier 3 K&A Statement: Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Justification:

A. incorrect - this may normally occur as flow drops followed by flux dropping.

Candidates may select this if they do nCit correctly recall parameters and logic needed to assess the status of THI.

B. Correct per ON step 3.3.3.a: "Peak to peak oscillations trending towards 10%

on APRMs (Oscillations measured from minimum peak to maximum peak)

C. incorrect - LPRM indications for oscillations have a 1-5 second period. Candidates may select this if they do not correctly recall that LPRM alarm frequency of 1-5 seconds is required.

D. incorrect - core flow has some cycling due to flow noise or may indicate a Recirc problem. Candidates may select this if tlhey do not correctly recall parameters and logic needed to assess the status of THI.

KIA Match Justification:

This question matches the stated KIA since candidates must recall the parameters and logic used to assess the presence of thermal hydraulic instabilities and the potential for fuel damage.

References:

ON-178-002 rev 16 Reference Required none Learning Objective: 15308 Question source: INPO Bank #23870 Question History: SSES 2002 NRC Exam Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41 (b)7 Comments: Created/Modified by: Bank Reviewed by: T. Ebert, L.. Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 73 Which one of the following conditions will require entry into EO-1 00-1 02, "RPV Control", EO-1 00-1 03, "PC Control", AND EO-100-113, "Level/Power Control".

(consider only current values of the stated parameters)

A. A small LOCA causes drywell pressure to rise to 1.80 psig, one control rod sticks at position 48, all other rods fully insert.

8. HPCI operation causes suppression pool water temperature to risato 108°F, and a manual scram results in NO rod motion.

C. A Main Turbine trip occurs, reactor power remains at 35%, and a loss of drywell cooling causes drywell temperature to rise to 145°F.

D. A loss of feed causes RPV water level to drop to +1", 10 control rods stick at position 48, and RCIC operation causes suppression pool level to rise to 23.5 ft.

LOC*23 NRC Exam Rev 3

K&A # 2.4.4 Importance Rating 4.5 QUESTION 73 RO Tier 3 K&A Statement: Ability to recognize abnormal indications for system operating parameters which arl9 entry-level conditions for emergency and abnormal operating procedures.

Justification:

A. Incorrect, conditions requiring UP control entry are not met with only one rod not full in. Candidates may select this if they do not correctly recall the entry conditions for UP control EOP.

B. Correct, SP temp >90F, and ATWS will require entry into all 3 of the stated EOPs C. Incorrect, drywell temp is not high enough for PC control entry. Candidates may select this if they do not correctly recall the entry conditions for PC control EOP.

D. Incorrect, SP level is not high enough for PC Control entry. Candidates may select this if they do not correctly recall the entry conditions for PC control EOP.

KA Match Justification:

This question matches the stated KIA since candidates must correctly recognize abnormal parameter indications that will require EOP entry.

References:

EO-000-102 rev 8, -103 rev 7, -113 rev Reference Required none 8

Learning Objective: 14585 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)10 Comments: Created by: T. North, 6/22110 Reviewed by: T. Ebert, L Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 74 Which one of the following will result in a violation of the Unit 1 facility operating license, NPF-14?

A. Continued operation above 94.4% Core Thermal Power.

B. Operation at power with NO reactor recirc pumps in operation.

C. Briefly exceeding 100.1 % of 3952 MWth on NBA01, "CTP Instantaneous".

D. Operation at power within region 1 of TRM 3.2.1 "Power/Flow Map".

LOC-23 NRC Exam Rev 3

K&A# 2.2.38 Importance Rating 3.6 QUESTION 74 RO Tier 3 K&A Statement: Knowledge of conditions and limitations in the facility license.

Justification:

A. Incorrect, 94.4% is the current license limit for unit 2 and was the previous limit for unit 1. Candidates may select this if they do not correctly recall unit 1 license power limitations.

B. Correct, Power operation with natural circulation is forbidden by Unit 1 Facility Operating License NPF-14 C. Incorrect, briefly exceeding 100.1 % of the license thermal power limit is expected and allowed lAW ON-100-004, "Reactor Power Greater Than Authorized Limif'.

Candidates may select this if they incorrectly believe that exceeding the license power limit for any amount of time is not permitted.

D. Incorrect, operation in region 1 is not restricted by the operating license. Candidates may select this if they do not correctly recall license limits.

KIA Match Justification:

This question matches the stated KIA since candidates must recognize a situation that is prohibited by the facility operating license.

References:

ON-100-004 rev 14, Unit 1 Facility Reference Required none Operating License NPF-14 Learning Objective: LP017 obj 4.a; 15299 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 41(b)7 Comments: Created by: T. North, 11/22110 Reviewed by:

LOC-23 NRC Exam Rev 3

QUESTION 75 Unit 1 was operating at 100% power when a small steam leak occurs inside the Reactor Building steam tunnel.

• Main Steam tunnel temperature is 1eO°F, up slow

• Main Steam Isolation Valves (MSIV) are OPEN

• All other Reactor Building temperatures are normal Which one of the following sets of actions must the crew take as a DIRECT result of the trend in steam tunnel temperature?

A.

• Enter ON-159-002, "Containment Isolation", AND EO-100-104, "Secondary Containment Control"

• Verify that the MSIVs shut after Cl 15 minute time delay B.

• Enter ON-159-002, "Containment Isolation", ONLY

• Verify that the MSIVs shut after a 15 minute time delay C.

• Enter ON-159-002, "Containment Isolation", AI\lD EO-100-104, "Secondary Containment Control"

• Shut the MSIVs D.

• Enter ON-159-002, "Containment Isolation", ONLY

• Shut the MSIVs LOC*23 NRC Exam Rev 3

K&A# 2.4.2 Importance Rating 4.5 QUESTION 75 RO Tier 3 K&A Statement: Knowledge of system set points, interlocks and automatic actions associated with EOP entry conditions.

Justification:

A. Incorrect, the MSIVs will isolate immediately upon exceeding 17]DF in the RB steam tunnel without a time delay. Candidates may select this if they believe the MSIVs have a 15 minute time delay that is common with other steam valve isolations.

B. Incorrect, entry conditions for SC Control EOP have been met arlOMSIVs should already be shut. Candidates may select this if they believe the MSIVs have a 15 minute time delay that is common with other steam valve isolations, and do not recognize the entry to the SC control EOP.

C. Correct, MSL tunnel temperature >157°F is an entry condition to SC Control EOP. The isolation setpoint of 177°F for MSIVs has also been exceeded requiring entry into ON-159-002. Since the MSIVs have failed to close, the crew should immediately complete the isolation manually.

D. Incorrect, entry conditions for SC Control EOP have been met. Candidates may select this if do not recognize the entry to the SC control EOP.

KIA Match Justification:

This question matches the stated KIA since candidates must recall knowledge of MSIV isolation setpoints that are associated with Secondary Containment control EOP entry.

References:

RM-OP-059B rev 5, EO-000-104 rev 6, Reference Required none ON-159-002 rev 29 Learning Objective: 14583.c Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 41 (b)7 Comments: Created by: T. North, 6/22110 Reviewed by: T. Ebert, L Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 76 SROONLV Unit 1 is at in MODE 1 at full power with the following conditions:

• Total core flow is 105 Mlbm/hr

• BOTH Recirc pump speeds are at 90%

• Recirc Pump A scoop tube positioner control power is de-energized Moments later a spurious Recirc Speed Limiter #1 runback is initiated.

Which one of the following is the status of the Reactor Recirc system; AND what action must be taken?

A.

• Pump A is at 90% speed, Pump B is at 30% speed; Operating in the allowable region of the Power/Flow map at 78% power.

• Declare loop B "out of service" due to flow mismatch per Tech Spec 3.4.1, "Recirculation Loops Operating".

B.

• Pump A is at 90% speed, Pump 13 is at 30% speed; Operating in the allowable region of the Power/Flow map at 78% power.

• Take manual control of the A scoop tube and reduce speed to 30%

per Tech Spec 3.4.1, "Recirculation Loops Operating".

C.

• BOTH pumps are at 30% speed; Operating in region 1 of the Power/Flow map at 63% power.

• Insert control rods to exit region 1 per ON-164-002, "Loss Of Reactor Recirculation Flow".

D.

• BOTH pumps are at 30% speed; Operating in region 1 of the Power/Flow map at 63% power.

• Reset Limiter #1 and raise Pump B speed to exit region 1 per 01\1 164-002, "Loss Of Reactor Recirculation Flow",

LOC-23 NRC Exam Rev 3

K&A # 295001 AA2.01 Importance Rating 3.7 QUESTION 76 SRO Tier 1 Group 1 K&A Statement: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Power/Flow Map Justification:

A. Correct, the A recirc pump will remain at 90% speed due to the loss of power to the scoop tube, and B will runback to 30%. This will not result in operation in the restricted regions of the p/f map. TS 3.4.1 will require declaring the lower flow loop out of service.

B. Incorrect, reducing speed of the A pump will result in intentional operation in region

1. candidates may select this if they incmrectly believe they must take manual action to match the recirc pump speeds. '

C. Incorrect, pump A will not run back. Candidates may select this if they incorrectly assume the speed limiter will act to reduce the A pump speed.

D. Incorrect, see C above.

KIA Match and SRO Only Justification:

This question matches the stated KIA since SRO candidates must evaluate the status of recirc flow and reactor power with respect to the power to flow map, and determine the correct action required.

References:

TRM 3.2 rev 11 ; TS 3.4.1 rev 3 Reference Required none Learning Objective:

Question source: Modified INPO Bank #27020 Question History: SSES 2003 NRC Exam Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)5 Comments: modified Modified by: T. North, 11-17-10 initial condition Reviewed by:

resulting in a different correct answer.

LOC*23 NRC Exam Rev 3

QUESTION 77 SRO ONLY Unit 1 has experienced an accident with a Primary System is discharging into Secondary Containment (SC) and significant fuel failure.

An uncontrolled offsite release is in progress, AND an ALERT EAL declaration has been declared due to offsite release rates ABOVE the ALERT level.

The crew should (1) because (2)

A. (1) isolate ALL Primary systems discharging into SC (2) it is the most direct and effective method for terminating th.e radioactivity release.

B. (1) isolate ALL Primary systems discharging into SC (2) actions required for the emergency plan have a higher priority than those required for EOPs.

C. (1) isolate ONLY those Primary systems discharging into SC that e NOT required to support EOP/DSP actions (2) isolation of those systems may result in a much larger uncontrolled release.

D. (1) isolate ONLY those Primary systems discharging into SC that are NOT required to support EOP/DSP actions (2) EOP required actions have a higher priority than those required for the emergency plan.

LOC*23 NRC Exam Rev 3

K&A# 295038 2.4.18 Importance Rating 4.0 QUESTION 77 SRO Tier 1 Group 1 K&A Statement: High Offsite Release Rate:

Emergency Procedures / Plan: Knowledge of the specific bases for EOPs.

Justification:

A. Incorrect, systems needed for EOP/DSP actions should not be isolated. Candidates may select this if they do not correctly recall the required action and its basis B. Incorrect, see A above.

C. Correct, EO-100-105, Rad Release, step RR-2 requires that systems needed for important EOP or DSP actions remain in service because isolation of those systems and not taking the required actions may result in a much larger release.

D. Incorrect, EOP actions do not necessarily have priority over EP actions. Candidates may select this if they do not correctly recall the basis for this EOP step KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must evaluate plant conditions, determine the correct EOP action required, and recall the basis for that action.

References:

EO-000-105 rev 3 Reference Required none Learning Objective: 14594 Question source: Modified INPO bank #2ei837 Question History: SSES 2003 NRC exam Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43{b)5 Comments: modified Created by: T. North, 11/17/10 to raise to SRO level, Reviewed by:

and better match the selected 1</A.

LOC-23 NRC Exam Rev 3

QUESTION 78 SROONLY Both Units are operating at full power when a fire is detected in the Unit 2 turbine building. The fire brigade is activated.

Several minutes later the following conditions occur:

• Simplex alarm FIRE DET 106_Z4 ALM, "Control Structure Outside Air Intake" actuates

• A SLIGHT smell of smoke is detectecl in the control room

• The fire brigade reports that the fire is still in progress, but under control Which one of the following is REQUIRED?

A. Direct the abandonment of the Control Room due to smoke intrusion per ON-100(200)-009, "Control Room Evacuation".

B. Direct actions to place the Smoke Removal System in service to ensure long term control room habitability per ON-013-001, "Response to Fire".

C. Direct actions to place the CREOASS system in PRESSURIZATION/FILTRATION MODE to isolate control room ventilation system from the source of smoke per ON-013-001 ,

"Response to Fire".

D. Direct actions to place the CREOASS system in RECIRCULATION MODE to prevent further smoke intrusion to the control room per ON 013-001, "Response to Fire".

LOC-23 NRC Exam Rev 3

K&A # 600000 AA2.03 Importance Rating 3.2 QUESTION 78 SRO Tier 1 Group 1 K&A Statement: Ability to determine and interpret the following as they apply to PLANT FIRE ON SITE: Fire alarm Justification:

A. Incorrect, conditions do not yet require control room evacuation. ON-100(200)-009 requires CR abandonment upon "denst smoke, extreme heat, or hazardous gas".

These conditions are not present, nor is there any other condition that should inspire Shift Supervision to abandon the control room. Candidates may select this if they believe the presence of the smell of smoke is sufficient to direct CR abandonment.

B. Incorrect, per ON-013-001 , the smoke removal system should not be placed in service until the fire is no longer in progress. Candidates may select this if they incorrectly believe the smoke removal system should be placed in service in this instance and do not correctly recall the mitigating strategies contained in ON-013 001 C. Incorrect, the pressurization mode of CREOASS, outside air intake to the control room is shifted to the CREOASS trains, and will not prevent smoke intrusion.

Candidates may select this if they do not correctly understand CREOASS system lineups or the mitigating strategies contained in ON-013-001.

D. Correct, ON-013-001 directs that CREOASS be placed in recirculation mode if smoke is detected in the MCR.

KIA Match Justification:

This question matches the stated KIA since candidates must interpret plant conditions following receipt of a fire alarm resulting from a fire on site.

SRO Only Justification:

This question is SRO only since candidates must correctly select and apply the required procedure and mitigating strategy based on plant conditions.

References:

ON-013-001 rev 28, TM-OP-030 rev 4, Reference Required none ON-100(200)-009 rev 21 , OP-030-002 rev 26 Learning Objective: 15306 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 43{b)5 Comments: Created by: T. North, 9/i/10 Reviewed by: T. Ebert, L Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUES1"ION 79 SRO ONLY Unit 1 was at full power when a reactor scram occurred.

The following conditions are currently present:

• RPV water level is being controlled a1t +20" with one reactor feed pump

• RPV pressure is being controlled with bypass valves in automatic

• 2 control rods are stuck at position 48

• All other control rods are at position 00

• Reactor power is midscale on IRM range 4, down slow

• Reactor period is -80 seconds

• 1\10 boron has been injected Which one of the following RPV pressure control strategies is CORRECT?

A. Commence a reactor cooldown < 1OO°F/hour UNLESS re-criticality is observed OR the shutdown cooling interlock clears.

B. Stabilize RPV pressure < 1087 psig UNTIL BOTH stuck rods are fully inserted, then commence a cooldown < 100°F/hour.

C. Stabilize RPV pressure < 1087 psig UI\ITIL AT LEAST 1 of the stuck rods is fully inserted, then commence a cooldown < 100°F/hour.

D. Stabilize RPV pressure < 1087 psig UNTIL Cold Shutdown Boron Weight has been injected, then commence a cooldown < 100°F/hour.

LOC-23 NRC Exam Rev 3

K&A # 295006 2.1.7 Importance Rating 4.7 QUESTION 79 SRO Tier 1 Group 1 K&A Statement: Scram:

Conduct of Operations: Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Justification:

A. Correct, per EO-OOO-113, LevellPower Control, if the reactor is subcritical on control rods and no boron has been injected, cooldown may commence

<100Flhr, unless re-criticality is observed.

B. Incorrect, cooldown may commence re~lardless of the status of the stuck rods as long as the reactor is subcritical and no boron has been injected. Candidates may select this if they incorrectly believe that cooldown cannot commence until the reactor meets the shutdown criteria for 13xiting the ATWS EOP, and do not correctly recall that criteria.

C. Incorrect, cool down may commence re~lardless of the status of the stuck rods as long as the reactor is subcritical and no boron has been injected. Candidates may select this if they incorrectly believe that cooldown cannot commence until the reactor meets the shutdown criteria for exiting the ATWS EOP.

D. Incorrect, injecting cold shutdown boron weight is not necessary and if boron injection is started cooldown will be precluded. Candidates may select this if they incorrectly determine that boron injection should be started and that they must wait for CSBW to commence a cooldown.

KIA Match Justification:

This question matches the stated KIA since candidates must evaluate IRM and SRM reactor power and period indications and determine that a reactor cooldown may commence in accordance with the EOP provided.

SRO Only Justification:

This question is SRO only since candidates must evaluate plant conditions and select the correct emergency procedural strategy based on that evaluation as required by 10 CFR 43(b)(5)

References:

EO-000-113-1 rev 9 Reference Required Learning Objective: 14622 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)5 Comments: Created by: T. North, 9(1/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 80 SROONLY Unit 1 has experienced a failure of the Electrohydraulic Control (EHC) system causing an uncontrolled RPV pressure rise.

The Reactor Protection System AND Alternate Rod Insertion systems failed to shutdown the reactor resulting in the following INITIAL transient conditions:

• RPV Pressure peaked at 1150 psig

• INITIALATWS power was 10%

SEVERAL MOMENTS LATER the following conditions are present:

• Reactor power is CURRENTLY Zo/.o,-OOwn slow due to boron injection

• Control rods have NOT yet been inserted

• RPV pressure is 1090 psig, being controlled with SRVs

• Suppression Pool temperature is 190°F, up slow

• Suppression Pool Level is 22 ft., up slow Given the Suppression Pool Temperature (SPIT) leg and figure 2, Heat Capacity Temperature Limit, from EO-1 00-1 03, "PC Control", determine which one of the following is the CORRECT action:

The Unit Supervisor must (1) because (2) .

A. (1) WAIT until the reactor is shutdown with control rods AND RPV pressure exceeds 1106 psig prior to directing Rapid Depressurization (2) large amplitude power swings may occur at low pressure and high power B. (1) WAIT until RPV pressure exceeds 1106 psig ONLY prior to directing Rapid Depressurization (2) the suppression pool can still absorb all the energy from the RPV without exceeding primary containment pressure limits.

C. (1) WAIT until the reactor is shutdown with control rods prior to directing Rapid Depressurization (2) large amplitude power swings may occur at low pressure and high power D. (1) DIRECT Rapid Depressurization NOW based on current plant conditions (2) the suppression pool may not absorb all the energy from the RPV without exceeding primary containment pressure limits.

LOC-23 NRC Exam Rev 3

K&A :fI: 295025 2.2.44 Importance Rating 4.4 QUESTION 80 SRO Tier 1 Group 1 K&A Statement: High Reactor Pressure:

Equipment Control: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Justification:

A. Incorrect, RD will only need to be postponed due to initial ATWS power. The current combination of RPV pressure, SP level and SP temp should be plotted on the unsafe side of the HCTL curve. RPV pressure rise above 1106 will not change this status.

Candidates may select this if they improperly interpret the HCTL curve.

B. Incorrect, since initial ATWS power was >5% RD must be postponed regardless of HCTL status due to potential power excursions at low RPV pressure. Current values of RPV pressure, SP level and temp should be plotted on the unsafe side of the HCTL curve, therefore, the SP currently MAY NOT be able to absorb RPV energy without exceeding 65 psig. Candidates may select this if they incorrectly evaluate the HCTL curve, and fail to recognize that RD must be postponed due to initial ATWS power.

C. Correct, per EO-00o-103, SPIT-5, if initial ATWS power is >5% further actions in ~

the SPIT leg may be postponed. Although the plot of RPV pressure, SP temp and level results in operation on the unsafe side of the HCTL curve, RD_must be postponed until the Rx is SID with control rods to preclude large power oscillations at low RPV pressure.

D. Incorrect, although operation is currently on the unsafe side of the HCTL curve and SP pool safety function is in jeopardy, RD cannot be performed at this time due to initial ATWS power >5%. Candidates may select this if they fail to recognize that RD is precluded by initial ATWS power.

KIA Match/SRO Only Justification:

This question matches the stated KIA since SRO candidates must evaluate and interpret plant indications and use these to evaluate the status of the suppression pool. They must further determine the correct EOP action to be directed and understand the impact this action will have on plant operation.

References:

EO-000-103 rev 7; EO-1 00-1 03 rev 9. Reference EO-1 00-1 03, SPIT leg Required and HCTL curve with SPOTMOS note removed, only.

Learning Objective: 14594 Question source: MODIFIED SSES OPS_INITIAL_LlCENSE bank

fI:PP002/2680 002 Question History: MODIFIED BANK Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)5 Comments: Created by: T. North, 9/9/10 Reviewed by: T. Ebert, L.. Casperson 11-12-10 LOC*23 NRC Exam Rev 3

QUESTION 81 SROONLY Unit 1 is operating at full power when I & C reports that LlS-B21-1 N025A, ATWS-RPT Level Indicating Switch, has failed.

Given Tech Spec 3.3.4.2, determine which one of the following is CORRECT for this situation:

A. Restore A TWS-RPT trip capability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> because this trip function is required to prevent jet pump and recirc pump cavitation when RPV water level is lowered in an A TWS event.

B. Restore the inoperable ATWS-RPT instrument channel within 14 days because this instrument is required to prevent jet pump and recirc pump cavitation when RPV water level is lowered in an ATWS event.

C. Restore ATWS-RPT trip capability within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> because this trip function is required to provide negative reactivity if a failure to scram event occurs.

D. Restore the inoperable ATWS-RPT instrument channel within 14 days because this instrument is required to provide negative reactivity if a failure to scram event occurs.

LOC-23 NRC Exam Rev 3

K&A # 295037 2.2.25 Importance Rating 4.2 QUESTION 81 SRO Tier 1 Group 1 K&A Statement: SCRAM Condition Present and Reactor Power Above APRM Downscale or unknown:

Eq.ylpment COl1trol: ~<nowledge of bases in technical specrrications for limlting conditions for operations and safety limits.

Justification:

A. Incorrect, ATWS-RPT trip capability is not lost since the B trip system is unaffected by the LIS failure, therefore the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action is not required. The basis for this function is not jet pump or recirc pump cavitation. Candidates may select this if they incorrectly determine trip capability is lost, and do not correctly recall the TS basis.

B. Incorrect, the basis for this function is not jet pump or recirc pump cavitation.

Candidates may select this if they do not correctly recall the TS basis.

C. Incorrect, ATWS-RPT trip capability is not lost since the B trip system is unaffected by the LIS failure, therefore the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> action is not required. Candidates may select this if they incorrectly determine trip capability is lost.

D. Correct, a failure of a single LIS in the ATWS-RPT logic makes one channel inoperable. Per TS 3.3.4.2 two channels are required per trip system. This function consists of 2 trip systems either of which can independently provide ATWS-RPT trip capability. Therefore, LCO condition A applies, and action A.1 is required ... restore the channel within 14 days. (A.2 may also be performed but this option is not addressed in this question)

The basis for this trip function is to provide negative reactivity by tripping recirc pumps in an ATWS event.

KIA Match Justification:

This question matches the stated KIA since candidates must correctly recall TS bases for instruments relied upon to mitigate an ATWS condition.

References:

TS & TSB 3.3.4.2 rev 0, TM-OP-080 Reference Required TS rev 9 3.3.4.2 Learning Objective: 10070 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)2 Comments: Created by: T. North, 7/11/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 82 SROONlY Unit 1 has experienced a seismic event resulting in a LOCA and an UN-ISOLABLE leak of the suppression pool into Secondary Containment: -------.. - -

• RPV Pressure is 850 psig, down slow

• Drywell pressure is 25.5 psig, up slow

• Drywell AND Suppression Chamber sprays are UNAVAILABLE

• Suppression Chamber pressure is 20 psig, up slow

• Suppression Pool water level is 18 feet, down slow Given figure 4 PSL below, determine which one of the following actions is required:

FIG4PSL f>RESSUFU! SUPf>RESSION LIMIT

~ 1Dl .. + + 1ft + . . . . . . .*. . ~. . + + +

28 .j.......... I * ,,******************..****..*+1* . .****** + ,., ;"""....1*********+ *1 ...... 26

~ 2Sr-4--+--r-~-+~~lj+--r~--T--r-+--r 26

~ ?*~-,--~*-4~-**~-+-*+*~~~~t*--~t--+~+-****~-r:

~ nr-4-*+~41- +-~ .. . . ~-~-+-~~--

"~ 10r-~-+--~~-+-r-~+--r-- ~o 18 A. Fully open ALL turbine bypass valves UNTIL suppression pool level drops to 14.5 ft, then perform EO-100-112, "Rapid Depressurization".

B. Perform EO-100-112, "Rapid Depressurization", NOW because the Pressure Suppression Limit WILL BE exceedEld.

C. Cooldown the RPV :S 100°F/hr UNTIL suppression pool level drops to 12 ft, then perform EO-100-112, "Rapid Depressurization".

D. Fully open ALL turbine bypass valves UNTIL suppression chamber pressure reaches 22 psig, then perform EO-100-112, "Rapid Depressurization".

LOC-23 NRC Exam Rev 3

K&A # 295030 2.4.6 Importance Rating 4.7 QUESTION 82 SRO Tier 1 Group 1 K&A Statement: Low Suppression Pool Water Level:

Emergency Procedures/Plan: Knowledge of EOP mitigation strategies.

Justification:

A. Incorrect, PSL curve status requires RD since parameters cannot be maintained in the safe region of the curve. Candidates may select this if they believe they must wait until SP level reaches unsafe side of the curve before RD is performed.

B. Correct, the rapidly lowering SP level and rising SC pressure will soon result in operation in the unsafe region of the PSL curve, therefore parameters cannot be maintained in the safe region. This requires RD now.

C. Incorrect, PC parameters cannot be maintained on the safe side of figure 4, therefore RD is required now. Candidates may select this if they believe they must wait until they reach the unsafe region clue to SP level reaching 12'.

D. Incorrect, PC parameters cannot be maintained on the safe side of figure 4, therefore RD is required now. Candidates may select this if they believe they must wait until they reach the unsafe region clue to SC pressure.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must evaluate plant conditions and correctly determine the required EOP mitigating strategy.

References:

EO-000-103 rev 7; EO-1 00-1 03 rev 9, Reference Required none PP002, rev 10 Learning Objective: 14622, 14624 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)2 Comments: Created by: T. North, 11/16/10 Reviewed by:

LOC-23 NRC Exam Rev 3

QUESTION 83 SROONLY Unit 1 is operating at 100% power with Control Rod Drive (CRD) Pump "B" out of service for corrective maintenance.

Unit 2 is in mode 4 with the CRD Hydraulic system out of service for a major modification.

Unit 1 CRD Pump "A" then trips due to a lockout and cannot be restarted. Electrical maintenance reports that the pump motor has failed.

The following is a timeline of events:

Time Event (1TIinutes)

Unit 1 CRD pump "A" trips I To FIRST accumulator trouble alarm To + 10 NPO reports HCU 22-23 pressure is 935 psig, down slow:

• To + 13 Rod 22-23 is at position 24 and is declared INOPERAB~E SECOND accumulator trouble alarm To + 15 NPO reports HCU 42-15 pressure is 940 psig, down slow.

To+ 18 Rod 42-15 is at position 48 and is declared INOPERABLE Which one of the following is the REQUIRED action?

A. Enter GO-1 00-004, "Plant Shutdown to Minimum Power" and, '

COMMENCE A REACTOR SHUTDO'NN IMMEDIATELY.

B. Enter ON-100-001, "Scram, Scram Imminent" and place the reactor MODE SWITCH TO SHUTDOWN IMMEDIATELY.

C. Enter ON-100-001, "Scram, Scram Imminent" and place the reactor MODE SWITCH TO SHUTDOWN PRIOR TO time To + 35 D. Enter ON-1 00-001, "Scram, Scram Imminent" and place the reactor MODE SWITCH TO SHUTDOWN PRIOR TO time To + 38 LOC-23 NRC Exam Rev 3

K&A # 295022 AA2.02 Importance Rating 3.4 QUESTION 83 SRO Tier 1 Group 2 K&A Statement: Ability to determine and/or interpret the following as they apply to LOSS OF CRD PUMPS: CRD system status Justification:

A. Incorrect, the normal shutdown procedure will not meet the procedural requirement to place the Rx mode switch to SID within 20 minutes. Candidates may select this if they choose the incorrect procedure to comply with the 20 minute requirement.

B. Incorrect, The immediate scram requirement is not applicable unless RPV pressure is <900#, RPV pressure at 100% power is normally >1000#. Candidates'may select this if they incorrectly apply the immediate scram requirement with RPV pressure

>900#

C, Incorrect, the scram REQUIREMENT will not expire until T +38. Candidates may select this if they incorrectly apply the 20 minute requirement to the receipt of the 2nd HCU low pressure alarm at T+ 15. This alarm comes in at 975, so the accumulator should not be declared inop until confirmed <940 psig.

D. Correct, ON-155-007, Loss of All CRD Flow, requires that the mode switch be placed in shutdown within 20 minutes following the DISCOVERY of the 2nd inop control rod due to low accumulator pressure. HCU 42*15 was discovered to be <940 psig at T+18, therefore the scram is REQUIRED prior to T+38.

Cross connect to Unit 2 CRD system is unavailable due to U2 status.

KIA Match Justification:

This question matches the stated KIA since candidates must interpret the status of HCU operability following a loss of both CRD pumps.

SRO Only Justification:

This question is SRO only since SRO candidates must evaluate system status and determine the required, procedurally directed mitigating strategy following the declaration of multiple inoperable control rods.

References:

ON-155-007 rev 21; TM-OP-055, rev 5 Reference Required none Learning Objective: 10034.c Question source: SSES OP002 Requal Bank

1. AD045/15304 008 Question History: Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)5 Comments: Created/Modified by: Bank Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 84 SROONLY Unit 1 is operating at full power when a central control rod drifts from position 24 to position 48.

The STA reports that the 60 minute average Core Thermal Power is 3954 MWth.

The Unit Supervisor (US) enters and directs actions in accordance with ON-155 001, "Control Rod Problems" due to the drifting control rod.

Which ADDITIONAL procedure(s) must the US enter, AND what action must be taken?

A.

• ON-100-004 "Reactor Power Greater Than Authorized Limit" AND ON 156-001, "Unanticipated Reactivity Change"

• Attempt to select and insert the drifting control rod then reduce pow~r with recirc flow if the rod will not remain at "00".

B.

• ON-100-004 "Reactor Power Greater Than Authorized Limit", AND ON-156-001, "Unanticipated Reactivity Change"

• Individually scram the drifting control rod then disarm the HCU.'

C.

• ON-156-001, "Unanticipated Reactivity Change" ONLY

• Declare the drifting control rod inoperable, insert it to "00", then disarm the HCU.

D.

• ON-100-004 "Reactor Power Greater Than Authorized Limit" ONLY

• Attempt to select and insert the drifting control rod then reduce power with recirc flow if the rod will not remain at "00",

LOC-23 NRC Exam Rev 3

K&A # 295014 AA2.01 Importance Rating 4.2 QUESTION 84 SRO Tier 1 Group 2 K&A Statement: Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION: Reactor Power Justification:

A. Correct, ON-100-004 requires entry since CTP has exceeded 100% of the 60 minute average of 3952 MWth. ON-156-001 requires entry due to the unanticipated positive reactivity insertion due to the drifting rod. ON-155-001 requires that the rod be selected and driven in. if the rod will not remain at 00, '

power must be reduced using recirc flow.

B. Incorrect, individual scram is not directed by procedure.

C. Incorrect, ON-100-004 also requires entry since the 60 minute average limit of 3952 MWth has been exceeded. Candidates may select this if they do not correctly. / '

evaluate the status of license power limits. ~ "

D. Incorrect, ON-156-001 should also be entered since an unanticipated reactivity addition has occurred. Candidates may select this if they do not correctly identify all required procedure entries.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must interpret the current value of reactor power and determine the correct procedure and required action action to be taken.

References:

ON-100-004 rev 16; ON-155-00*1 rev Reference Required none 35; ON-156-001 rev 22 Learning Objective: 15306 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)5 Comments: Created by: T. North, 9/'15/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 85 SRO ONLY Unit 1 has experienced a PRIMARY system leak into SECONDARY Containment.

EO-100-102, "RPV Control", and EO-100-104, "Secondary Containment Control" have been entered.

Reactor Building 749' Fire Suppression System Simplex Fire Alarm X218_Z7 is ALARMING.

Which one of the following is the operational impact of this alarm; AND what action should the crew take?

A.

• Reactor building temperatures are rising toward max safe values.

• Take actions to anticipate rapid depressurization.

B.

• Reactor building 749' level may be inaccessible due to a potential fire.

• Immediately activate the Fire Brigade.

C.

• Reactor building 749' level may become flooded due to fire system initiation.

• Evaluate further EOP actions for Secondary Containment flooding D.

• Reactor building 749' level may be above 212°F

• Limit the use of the RPV Wide Range Level indicator to ABOVE -125".

LOC-23 NRC Exam Rev 3

K&A# 2950322.4.20 Importance Rating 4.3 QUESTION 85 SRO Tier 1 Group 2 K&A Statement: High Secondary Containment Temperature:

Emergency Procedures / Plan: Knowledge of the operational implications of EOP cautions, warnings or notes Justification:

A. Incorrect, this alarm is not used for the purpose of evaluating RB temps in preparation for RD. Candidates may select this if thE~y do not recall the proper use of this alarm for EOP actions.

B. Incorrect, while this alarm may indicate that a fire may be present, the fire alarm should not be activated until confirmation is received. Candidates may select this if they do not recall the proper use of this alarm for EOP actions.

C. Incorrect, this alarm is not used for the purpose of evaluating flooding conditions in the RB. Candidates may select this if they do not recall the proper use of this alarm for EOP actions.

D. Correct. Per EO-OOO-100 caution 1, this fire alarm provides indication that the area near the wide range level instrument rack may be above 212F. this requires that the wide range level indicator not be used below *125" KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must recall knowledge of EOP caution #1 and determine the operational implications of reaching those conditions.

References:

EO-000-100, rev 5; EO-1 00-1 04, rev 6

References:

Learning Objective:

Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 43(b)5 10CFR55 Comments: Created by: T. North, 9/~~8/10 Reviewed by: T. Ebert, L.. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 86 SROONLY With Unit 1 at full power, Recirc Flow Transmitter FT-B31-1 N014A fails DOWNSCALE.

Which one of the following describes the expected response of the Power Range Neutron Monitoring System (PRNMS); AND what procedure should the Unit Supervisor implement?

A.

• Half-scram from the two out of four voters in Division I for APRM 1; Rod Block; "APRM Flow Reference Off-Normal" Alarm AR-103-E06;

• GO-1 00-012, "Power Maneuvers", to reduce power and clear the Half scram, Rod Block, and APRM Flow Reference Off-Normal Alarm.

B.

• Half-scram from the two out of four voters in Division I for APRM 1; Rod Block; "APRM Flow Reference Off-Normal" Alarm AR-103-E06;

• AR-1 03-001-E06 to bypass APRM 1 and clear the Rod Block and Half-scram.

C.

• Single Vote on all two out of four voters for APRM 1; Rod Block; "APRM Flow Reference Off-Normal" Alarm AR-103-E06;

• AR-1 03-001-E06 to bypass APRM 1 and clear the Rod Block and APRM vote.

D.

• Single Vote on all two out of four voters for APRM 1; Rod Block; "APRM Flow Reference Off-Normal" Alarm AR-1 03-E06;

• GO-100-012, "Power Maneuvers", to reduce power and clear the Rod Block, APRM vote, and APRM Flow Reference Off-Normal Alarm.

LOC-23 NRC Exam Rev 3

K&A # 215005 A2.05 Importance Rating 3.6 QUESTION 86 SRO Tier 2 Group 1 K&A Statement: Ability to (a) predict the impacts of the following on the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions Loss of recirculation flow signal Justification:

A. Incorrect, No half scram will occur. Power reduction is not necessary, since actual power has not changed. Candidates may select this if they believe conditions will generate a half scram and that a power reduction is necessary to clear the condition.

B. Incorrect, no half scram will occur. Candidates may select this if they believe conditions will generate a half scram.

C. Correct, with flow sensed by APRM 1 now significantly reduced, power will be above the rod block and flow biased trip setpoints. This generates a single APRM vote and a rod block. The APRM off normal flow alarm will actuate due to a >7% flow comparator signal.

The alarm response procedure provides direction for bypassing APRM 1. This will clear the input to RPS and RMCS.

Whit a failed flow transmitter, the off normal flow alarm will remain lit.

D. Incorrect, no power reduction is necessary. Candidates may select this if they believe a power reduction is necessary to clear the condition.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must predict the plant impact of a failure of the recirc flow transmitter, and select the correct procedural strategy to correct the condition.

References:

TM-OP-078D rev 6, AR-103-001 E06 Reference Required none rev 38 Learning Objective: 15716 Question source: Modified SSES OP002 Hequal Bank #TMOP078D/15716 003 Question History: Modified Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)5 Comments: Modified by: T. North, 7/19/10 Reviewed by: T. Ebert, l.. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 87 SROONlY Unit 1 was operating at full power when an ATWS occurred.

The following conditions are present:

• Initial ATWS power was 35%, and is currently 25%, down slow

• MSIVs are open

• SBLC pump A injecting at 30 GPM

• SBLC pump B is unavailable Given the condition of the SBLC system, the crew should ...

A. secure SBLC injection AND inject boron with RCIC because Suppression Pool Heat Capacity Temperature Limits may be exceeded.

B. continue to inject boron with SBLC ONLY because cold shutdown boron weight can still be achieved.

C. continue to inject boron with SBLC AND inject boron with RCIC because cold shutdown boron weight may not be achieved using SBLC.

D. Secure SBLC injection AND inject boron with RCIC because cold shutdown boron weight cannot be achieved using SBLC.

LOC-23 NRC Exam Rev 3

K&A # 211000 A2.04 Importance Rating 3.4 QUESTION 87 SRO Tier 2 Group 1 K&A Statement: Ability to (a) predict the impacts of the following on the STANDBY LIQUID CONTROL SYSTEM; and {b} based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Abnormal System Flow Justification:

A. Incorrect, RCIC should not be used in this situation since SBlC injection at reduced flow will still result in significant power reduction and ultimately CSBW. RCIC is used only if SBlC pumps are unavailable, since RCIC is lined up to utilize the SBlC tank.

HCTl is not directly related to the minimum flow requirements for SBlC. Candidates may select this if they believe the reduced SBlC flow warrants utilizing RCIC for boron injection and incorrectly believe the reduced flow will threaten HCTL.

B. Correct, although SBlC flow is below the expected flowrate for one pump, the reduced flowrate will not prevent CSBW from being achieved. CSBW is not related to the rate of boron injection.

C. Incorrect, RCIC cannot be used in conjunction with SBlC. Candidates may select this if they believe the reduced SBlC flow warrants utilizing RCIC for boron injection and that it can be used in conjunction w~th SBlC.

D. Incorrect, CSBW can be achieved using SBlC. Candidates may select this if they believe the reduced 'flow rate will prevent achievement of CSBW.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO Candidates must predict the impact of abnormal SBlC system flow and select the correct procedurally directed mitigating strategy to implement.

References:

EO-000-113 rev 8, ES-150-002 rev 19 Reference Required none learning Objective: 1214.g Question source: MODIFIED SSES OPS_INITIAl_LlCENSE Bank

1. TMOP053/1205001 Question History: Modified Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b}2 Comments: modified by: T. North, 9/20/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 88 SROONlY Unit 1 is in Mode 4 preparing for a plant startup when a review of In Service Testing (1ST) program records indicate that the safety function lift setpoint was UNSATISFACTORY for three (3) Safety Relief Valves (S/RV).

All other S/RV's setpoints are satisfactory.

Given Tech Spec 3.4.3, determine which one of the following is correct and the reason why:

Unit 1 CANNOT transition to Mode 2 until:

A. ALL THREE (3) of the S/RVs have been repaired because they are required to prevent the reactor vessel from exceeding its design pressure of 1250 psig.

B. AT LEAST ONE (1) of the S/RVs has been repaired because they are required to prevent the reactor vessel from exceeding its design pressure of 1250 psig.

C. ALL THREE (3) of the S/RVs have been repaired because they are required to prevent the reactor vessel from exceeding its ASME code pressure limit of 1375 psig.

D. AT LEAST ONE (1) of the S/RVs has been repaired because they are required to prevent the reactor vessE~1 from exceeding its ASME code pressure limit of 1375 psig.

LOC-23 NRC Exam Rev 3

K&A # 239002 2.2.25 Importance Rating 4.2 QUESTION 88 SRO Tier 2 Group 1 K&A Statement: Safety Relief Valves:

Equipment Control: Knowledge of the bases in Technical Specifications for limiiting conditions for operations and safety limits Justification:

A. Incorrect, the LCO does not require all 3 to be repaired, since 13 SRVs are currently operable, only one needs to be repaired. SRVs do NOT prevent exceeding the RPV design pressure. Candidates may select this if they do not correctly apply TS required actions, nor correctly recall TS basis for SRV operability.

B. Incorrect, SRVs do NOT prevent exceeding the RPV design pressure. Candidates may select this if they cannot correctly recall the correct TS basis.

C. Incorrect, the LCO does not require all 3 to be repaired, since 13 SRVs are currently operable, only one needs to be repaired. Candidates may select this if they incorrectly apply TS requirements.

D. Correct, TS 3.4.3 requires that the safety function of 14 SRVs are operable, or the plant must be in mode 4. Repair to one SRV will satisfy the LCO to enable transition to mode 2. The purpose of the safety function is to prevent the RPV from exceeding the ASME code limit of 1375 psig following analyzed pressure transients.

KIA Match Justification:

This question matches the stated KIA since candidates are required to recall the tech spec basis for SRV operability.

SRO Only Justification:

This question is SRO only since candidates must determine the actions required to satisfy tech specs for SRV operability and changing modes, and recall the basis for this tech spec.

References:

TS 3.4.3 & bases rev 2 Reference Required TS 3.4.3 Learning Objective: 1655.a, 13400 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)2 Comments: Created by: T. North, 7/12/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUES"rlON 89 SRO ONLY Unit 1 is operating at full power with the following conditions present:

• "A" TBCCW Pump and Heat Exchanger are in service

• AR-123-GOS, "TBCCW HEADER HI-LO TEMP" is received

• TBCCW Cooler Temp TIC-10946 vertical meter is FAILED DOWNSCALE

• TIC-10946 will NOT respond in MANUAL The Unit Supervisor should enter ON-11S-001, "Loss of TBCCW", and direct the ...

A. NPO to throttle OPEN the TBCCW HX Temp CV Bypass Valve (BPV 101083) to LOWER TBCCW header temperature.

B. NPO to MANUALLY throttle CLOSED the rBCCW HX Temp Control Valve (TV 10946) to RAISE TBCCV\l header temperature.

C. PCOM to place the "B" TBCCW heat exchanger in service to LOWER rBCCW header temperature.

D. PCOM to align emergency service water "A" TBCCW heat exchanger to LOWER TBCCW header temperature.

LOC*23 NRC Exam Rev 3

K&A # 400000 2.4.11 Importance Rating 4.0 QUESTION 89 SRO Tier 2 Group 1 K&A Statement: Component Cooling Water System:

Emergency Procedures / Plan: Knowledge of abnormal condition procedures.

Justification:

A. Correct, the TIC false downscale failure provides input to the TIC to attempt to raise TBCCW header temp by throttling closed TV 10946 to reduce SW flow.

This results in high TBCCW temp. ON-115-001 requires that the temperature control valve bypass be manually throttled open locally by NPOs to raise SW flow and reduce TBCCW temp.

B. Incorrect, the HI-LO alarm is due to high temp resulting from the TIC failure. The bypass valve should be throttled open to provide additional SW flow to lower TBCCW temps. Candidates may select this if they incorrectly diagnose the result of the TIC failure.

C. Incorrect, since the TIC and SW control valve are common to both A and B HXs, this will not have the desired effect. Candidates may select this if they do not recall this fact.

D. Incorrect, this action would only be taken if the CV bypass valve cannot be manually opened. Candidates may select this if they do not correctly recall the proper mitigating strategy sequence.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must recall knowledge abnormal condition procedure ON-115-001 strategy required to mitigate a degradation of the TBCCW system.

References:

ON-115-001 rev 17, TM-OP-01ei rev 4 Reference Required none Learning Objective: 15304 Question source: SSES OPS_INITIAL_LlCENSE Bank #AD045/15304 020 Question History: Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43{b)5 Comments: Created/Modified by: Bank Reviewed by: T. Ebert, l.. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 90 SROONlY Unit 1 is operating at full power with Emergency Diesel Generator "A" inoperable. The DG has been disassembled for repairs to the cylinder liners, and the repair is expected to take 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

LCO 3.8.1 condition B is in effect, and all required actions have been taken.

Electrical engineering later reports that Startup Transformer T-10 must be declared inoperable due to oil contamination discovered during a recent sample.

Repairs are expected to take a minimum of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

The following is a timeline of events:

0600 DG "A" declared inoperable Day 1 1200 DG "A" repairs commence I no  ? 0600 T-10 declared inoperable 1--" 0800 T-10 repairs commence Given Tech Spec 3.8.1 determine which one of the following identifies the required actions?

A. Be in mode 3 by day 2 at 1800, and mode 4 by day 3 at 1800.

B. Be in mode 3 by day 3 at 0600, and mode 4 by day 4 at 0600.

C. Be in mode 3 by day 3 at 0800, and mode 4 by day 4 at 0800.

D. Enter LCO 3.0.3 on day 2 at 0600, and be in mode 3 by day 2 at 1900.

LOC-23 NRC Exam Rev 3

K&A # 264000 2.2.36 Importance Rating 4.2 QUESTION 90 SRO Tier 2 Group 1 K&A Statement: Emergency Generators (Diesel/Jet):

Equipment Control: Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Justification:

A. Incorrect, condition 0 allows 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> to repair either the DG or T -10, and condition F is not applicable until that time. Candidates may select this if they incorrectly determine that condition F applies.

B. Correct, TS 3.8.1.0 applies since T-10 is one of the 2 offsite circuits, and requires restoration of the OG or T-10 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> from the time T-10 is declared inop {day 2 0600). If the repair times are accurate, then condition F will apply (on day 2 at 1800) and mode 3 is required 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> following on day 3 at 0600. Mode 4 is required 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> from meeting condition F, on day 4 at 0600.

C. Incorrect, the required time starts when T-l0 is declared inop, not when maintenance commences. Candidates may select this if they incorrectly determine when the tech spec clock starts.

D. Incorrect, this is only applicable if condition G is met, and it is not. Candidates may select this if they incorrectly determine that condition G is applicable.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must evaluate and apply tech spec LCD required actions resulting from diesel generator maintenance activities.

References:

TS 3.8.1 rev 4, TSB 3.8.1 rev 6 Reference Required TS 3.8.1 Learning Objective: 12555 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)2 Comments: Created by: T. North, 7/14/10 Reviewed by: T. Ebert, L.. Casperson 11-12-1 0 LOC*23 NRC Exam Rev 3

QUESTION 91 SROONLY Unit 1 is in mode 2 conducting a reactor startup with the following conditions present:

• Reactor Power is midscale on IRM range 6, steady

• Control rods are being withdrawn to maintain heatup rate Power is then lost to the plant process computer (PICSY) resulting in a loss of the Rod Worth Minimizer (RWM).

Given TS 3.3.2.1, determine which one of the following is correct:

(Consider ONLY the effect of the loss of the RWM)

The startup ...

A. CAI\II\JOT continue until power to PICSY and the RWM is restored because Tech Specs require that rod movement be immediately suspended.

B. CANNOT continue because the crew will NOT be able to verify that ~ 12 control rods have been withdrawn with the RWM unavailable.

C. CAN continue provided that the crew verify that a startup without the RWM has NOT been performed in the last calendar year; AND ensures that ~ 12 control rods have been withdrawn.

D. CAI\I continue provided that the crew verify that a startup without the RWM has I\IOT been performed in the last calendar year, AND ensures compliance with the pull sequence by a 2nd licensed operator.

LOC-23 NRC Exam Rev 3

K&A # 201006 A2.01 Importance Rating 2.8 QUESTION 91 SRO Tier 2 Group 2 K&A Statement: Ability to (a) predict the impacts of the following on the ROD WORTH MINIMIZER SYSTEM (RWM) (PLANT SPECIFIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Power supply loss: P-Spec(Not-BWR6)

Justification:

A. Incorrect, tech specs provides other methods to compensate for the loss of the RWM in order to continue the stu. Rod wId need not be suspended. Candidates may select this if they incorrectly believe that the RWM is required for the Stu to continue.

B. Incorrect, the 12 rod requirement is verifiable without the RWM. The stu may continue provided other required actions are met. Candidates may select this if they incorrectly believe that the RWM is required to verify 12 rods wId.

C. Incorrect, EITHER the 12 rod verification OR no previous stu w/o the RWM in a year must be performed. Additionally, a 2 nd operator/qualified person must verify the BPWS is followed. Candidates may select this if they incorrectly believe that both these requirements must be met.

D. Correct, startup procedure GO-100-002 note states that the startup may continue with the RWM inop provided TS 3.3.2.1 is met. This TS requires that a 2nd operator/qualified person AND EITHER: verify :::12 rods WID OR verify no S/U has been done w/o the RWM in the last calendar year.

KIA Match Justification:

This question matches the stated KIA since candidates must predict how the failure of the RWM will impact the startup, and apply a procedural note permitting continuation of the StU after compliance with TS.

References:

GO-100-002 rev 66, TS 3.3.2.1 rev 2 Reference Required TS 3.3.2.1 Learning Objective: 12567 Question source: Modified SSES OPS_INITIAL_LlCENSE Bank

1. TMOP031 0/12567 001 Question History: Modified Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)6 Comments: Modified by: T. North, 7/6/10 Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 92 SROONLY Unit 1 is operating at full power, when I&C reports that Reactor Steam Dome Pressure Switches, PS-B21-11\J023A AND B, have BOTH failed their quarterly surveillance test.

These switches provide the High Reactor Steam Dome Pressure Scram signal.

• PS-B21-11\J023A "as left" setpoint was 1093 psig

• PS-B21-11\J023B "as left" setpoint was 1097 psig NEITHER switch can be adjusted and BOTH must be replaced.

Suitable replacement pressure switches will NOT be available for 3 weeks.

Given Tech Spec 3.3.1.1, "RPS Instrumenta.tion" determine w~lich one of the following is REQUIRED:

A. Place channel A OR B in TRIP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; AND Be in mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

B. Place BOTH failed channels in TRIP within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

C. Be in mode 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ONLY.

D. Place channel A OR B in TRIP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> ONLY.

LOC-23 NRC Exam Rev 3

K&A # 216000 2.2.40 Importance Rating 4.7 QUESTION 92 SRO Tier 2 Group 2 K&A Statement: Nuclear Boiler Instrumentation:

Equipment Control: Ability to apply Tech Specs for a system.

Justification:

A. Incorrect, action G, mode 3 in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, will not apply provided action B can be performed. Candidates may select this if they incorrectly determine that the delay in repair will result in exceeding required action times.

B. Incorrect, condition A does not apply since one channel in each trip system is inoperable. Also, this action will result in a full scram. Candidates may select this if they incorrectly determine that condition A should be applied for each channel and attempt to apply tech specs without regard for the operational impact.

C. Incorrect, see A above.

D. Correct, action B applies, and requirt~s that 1 of the 2 inoperable channels be placed in TRIP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. No other action needs to be taken regardless of the extended replacement time.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates be able to apply tech specs following a loss of nuclear boiler instrumentation.

References:

TS 3.3.1.1 rev 3, SI-158-303 rev 22 Reference Required TS 3.3.1.1 with APRM setpoints redacted.

Learning Objective: 13426 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)2 Comments: Ensure Created by: T. North, 11/16/10 reference provided Reviewed by: E. Brice, A. Avery, 11-23-10 has APRM setpoints redacted to prevent providing answer to RO#7 for SROI candidates LOC-23 NRC Exam Rev 3

QUES1"ION 93 SROONLY Unit 1 is operating at 100% power when the Main Turbine trips due to false High Reactor Vessel Level signals.

After the Turbine trip, the Control Room Operators report the following conditions and alarms:

• Auxiliary Buses 11A (1A101) and 118 (1A102) transferred to Startup Bus 10 (OA103)

• Main Generator Sync breaker (1 R1 01) - OPEN

• 230 kV Switchyard Breakers, 3W (Generator 1 West) and 3T (Generator 1 East) - OPEN

• Main Generator Exciter Field Breaker - OPEN

• AR-106-A08, liGEN LOCKOUT RELAYS TRIP"

• AR-106-E08, liGEN ANTI MOTORING TRIP" What actions must be directed as a result of the above information?

A.

• EnterON-100-101, "Scram, Scram Imminent" and ON-003-001, "Loss Of Startup Bus 10"

• CONTACT Transmission Control Center (TCC) to investigate the cause of the 3W and 3T 230 KV breaker trip and reclose

• Re-energize Auxiliary Busses 11 A and 11 B B.

• Enter EO-1 00-1 02, "RPV Level Control" and ON-104-201, "Loss Of 4kv ESS Bus 1A & 1C"

• CONTACT Transmission Control Center (TCC) to re-energize Auxiliary Busses 11 A and 11 B

• Verify "A" & "C" DIGs running with cooling water C.

• Enter ON-1 00-1 01, "Scram, Scram Imminent and ON-193-002, "Main Turbine Trip"

• CONTACT the Scranton System Operator to investigate the cause of the 3W and 3T 230 kV breaker trip

• Restart Reactor Recirculation Pumps D.

• Enter ON-198-004, "Unit 1 Main Generator Unable To Disconnect From Grid After A Turbine Trip"

• CONTACT the Scranton System Operator to block open 230 kV breakers, 3W and 3T

• Verify AR-106-E08, "GEN ANTI MOTORING TRIP" cleared after 30 seconds LOC-23 NRC Exam Rev 3

K&A # 245000 A2.05 Importance Rating 3.8 QUESTION 93 SRO Tier 2 Group 2 K&A Statement: Ability to (a) predict the impacts of the following on the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS; and (b) based on those pred~ctions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Generator Trip Justification:

A. Incorrect, The startup bus has not been lost and the Aux buses are still energized.

Scranton not the TCC should be contacted to operate the switchyard breakers. If the candidate does not recognize this, this answer may be chosen.

B. Incorrect, Power was not lost to the Aux buses or the ESS buses. The power supplies to the aux buses have transferred, but power is automatically restored. If the candidate does not recognize this, this answer may be chosen.

C. Correct answer. ON-100-101 and ON-193-002 should be entered simultaneously. Scranton should be contacted to operate the switchyard breakers. The RR pumps tripped on EOC-RPT and should be restarted per procedure for forced circulation through the core.

D. Incorrect, The main generator has separated from the grid (the Main Generator Sync breaker is open). The operator should not enter ON-198-004. If the candidate does not recognize this, this answer may be chosen.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must predict the effect of the main turbine and main generator trip conditions and determine the correct procedural actions.

References:

ON-100-101 rev 25; ON-193-002 rev Reference Required None 17 Learning Objective: 15304 Question source: SSES NRC Exam Bank #127 Question History: SSES 2004 NRC Exam Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)5 Comments: Created by: Bank Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 94 SRO ONLY Due to a significant plant transient beyond the design basis of the plant, the Shift Manager has authorized an operation in accordance with 10CFR50.54X.

Which one of the following describes the action(s) required, if any?

NRC notification (1)  ; and NRC approval of the actions to be taken (2)

A. (1) IS NOT required (2) NOT required B. (1) IS required and MUST be made BEFORE OR IMMEDIATELY AFTER taking action (2) NOT required C. (1) IS required BEFORE taking action (2) IS required D. (1) IS required BEFORE taking action (2) NOT required LOC-23 NRC Exam Rev 3

K&A # 2.1.2 Importance Rating 4.4 QUESTION 94 SRO Tier 3 K&A Statement: Knowledge of operator responsibilities during all modes of plant operation.

Justification:

A. Incorrect, NRC notification is required. Candidates may select this if they are unfamiliar with SRO responsibilities regarding 10CFR50.54x notification requirements.

8. Correct, per OP-AD-001 NRC notification of 10CFR50.54X actions should be made prior to if practical, or immediately after action has been taken. NRC approval of the action is not required.

C. Incorrect, NRC notification can be made immediately after taking action, and NRC approval is not required. Candidates may select this if they are unfamiliar with SRO responsibilities regarding 10CFR50.54x notification requirements.

D. Incorrect, NRC notification can be made immediately after taking action. Candidates may select this if they are unfamiliar with SRO responsibilities regarding 10CFR50.54x notification requirements" KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must correctly recall their responsibilities with respect to compliance with 'IOCFR50.54X and OP-AD-001.

References:

OP-AD-001 rev 44 Reference Required none Learning Objective: 14715 Question source: SSES OPS_INITIAL_LlCENSE Bank # AD044!14715 002 Question History: Bank Cognitive level: Memory/Fundamental knowledge: X Comprehension!Analysis:

10CFR55 43(b)1 Comments: Created by: Bank Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 95 SRO ONLY Which of following examples would NOT require a Temporary Engineering Change?

A. A thermocouple is installed on the exterior of process pipe and connected to a recorder plugged into a wall outlet via a procedure or an open PCWO.

B. Installation of Watt-meters into the 13.8 kV System that monitor the output voltage and have the potential if they failed to adversely impact the 13.8 kV System.

C. A breaker is opened to isolate a component. The breaker also feeds components in other subsystems, which remain operable. There is no open WOo D. Isolate a circuit to a defective pump vibration probe to eliminate a nuisance alarm, or to eliminate a condition that is masking other valid alarms.

LOC-23 NRC Exam Rev 3

K&A# 2.2.5 Importance Rating 3.2 QUESTION 95 SROTier3 K&A Statement: Knowledge of the process for making design or operating changes to the facility.

Justification:

A. Correct - Temporary Engineering Change not required if the installation of temporary test instruments does not affect plant equipment nor have any potential to affect plant equipment.

B. Incorrect - Temporary Engineering Change may be required because this action is invasive. Candidates may select this if they do not correctly evaluate the need for a temporary engineering change.

C. Incorrect - Temporary Engineering Change may be required if components are not restored following the completion of troubleshooting activity and the WO is closed. Candidates may select this if they do not correctly evaluate the need for a temporary engineering change.

D. Incorrect - Temporary Engineering Change not required if the installation of temporary test instruments does not affect plant equipment nor have any potential to affect plant equipment. Candidates may select this if they do not correctly evaluate the need for a temporary engineering change.

KIA Match and SRO Only Justification:

This question matches the stated KIA since SRO candidates must determine whether a specific case meets the requirements of the process for implementing temporary engineering changes to the facility. Only SRO candidates are required to make this determination at SSES.

References:

NDAP-QA-1218 Reference Required none learning Objective: 15313 Question source: SSES OPS_INITIAl-LiCENSE Bank #AD044/15313 026 Question History: Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)2 Comments: Created by: Bank Reviewed by: T. Ebert, l. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUES"nON 96 SROONlY Which of the following sets of instruments are utilized during emergency plan implementation to evaluate the status of fission product barrier integrity per EP TP-001, "Emergency Classification Level Manual" Category F, "Fission Product Barrier Degradation"? .

A. Offgas Pre-treatment Logarithmic Rad Monitors and Main Steam Line Rad Monitors B. Turbine Building Area Rad Monitors and Containment High Range Monitors C. Turbine Building Area Rad Monitors and Reactor Building Area Rad Monitors D. Containment High Range Monitors and Reactor Building Area Rad Monitors LOC-23 NRC Exam Rev 3

K&A# 2.3.15 Importance Rating 3.1 QUESTION 96 SRO Tier 3 K&A Statement: Knowledge of radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc Justification:

A. Incorrect, neither of these are used for this purpose.

B. Incorrect, TB ARMs not used for this purpose.

C. Incorrect, TB ARMs not used for this purpose.

D. Correct, per EP-TP-001 table F, both the containment hi rad and RB ARMs can provide input to the determination of the status of fission product barriers.

KIA Match & SRO Only Justification:

This question matches the stated KIA for SRO only since candidates must recall knowledge of rad monitoring instrumentation utilized for determining the status of fission product barriers while evaluating EAL classification levels. This task is "SRO only" at SSES.

References:

EP-TP-001 rev 3 Reference Required none Learning Objective:

Question source: New Question History: New Cognitive level: MemorylFundamental knowledge: X ComprehensionlAnalysis:

10CFR55 43(b)4 Comments: Created by: T. North, 7/13/10 Reviewed by: T. Ebert, L. Casperson 11-12-1 0 LOC-23 NRC Exam Rev 3

QUESTION 97 SROONLY Unit 1 is conducting a reactor shutdown prior to a refueling outage per GO-1 00 004, "Shutdown To Minimum Power".

The reactor mode switch is placed in SHUDOWN as directed by the procedure.

As a result, reactor level drops to +8", and is subsequently restored to +35".

Given NDAP-QA-0720, "Station Report Matrix And Reportability Evaluation Guidance" Attachment K, determine which one of the following is REQUIRED:

A.

• Enter EO-1 00-102, "RPV Control" ONLY

• NO NRC notification is required B.

• Enter EO-1 00-1 02, "RPV Control" ONLY

• Make a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ENS notification to the NRC C.

• Enter EO-100-102, "RPV Control", AND ON-100-101, "Scram, Scram Imminent"

• NO NRC notification is required D.

• Enter EO-100-102, "RPVControl", AND ON-100-101, "Scram, Scram Imminenf'

• Make a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> ENS notification to the NRC LOC-23 NRC Exam Rev 3

K&A# 2.4.30 Importance Rating 4.1 QUESnON97 SRO Tier 3 K&A Statement: Emergency Procedures / Plan: Knowledge of events related to system operation/status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Justification:

A. Incorrect, ON-100-101 must be entered. Candidates may select this if they believe the scram procedure is not required.

B. Incorrect, ON-100-101 must be entered and no notification is necessary. Candidates may select this if they believe the scram procedure is not required, and incorrectly apply the NDAP.

C. Correct, RPV control must be entered because level dropped below +13". ON 100-101 entry is directed by the GO. No ENS notification is required since the scram is planned and expected by procedural direction.

D. Incorrect, no notification is required. Candidates may select this if they incorrectly apply the NDAP.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must.

References:

NDAP-QA-0720 attac~lment K rev 17; Reference Required NDAP EO-OOO-l 02 rev 9; GO-l00-004 rev 53 QA-0720 Att K Learning Objective: 14585 Question source: SSES OPS_INITIAL_L1CENSE Bank #PP002/14585 010 Question History: Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis:

10CFR55 43(b)5 Comments: Created by: Bank Reviewed by:

LOC-23 NRC Exam Rev 3

QUES"nON 98 SROONLY Which set of statements below correctly describes an EXAMPLE of an OPERABILITY TEST versus a FUNCTIONAL TEST in accordance with NDAP QA-0482, "Post-Maintenance/Modification Test Guidelines"?

A.

• An Operability Test demonstrates that the Service Water System can deliver water to the Main Generator Hydrogen Coolers.

• A Functional Test demonstrates RCIC can provide 600 GPM against a system head corresponding to Reactor pressure.

B.

• An Operability Test demonstrates that RCIC can provide 600 GPM against a system head corresponding to Reactor pressure.

• A Functional Test demonstrates that Service Water System can supply water to the Main Steam Pipe Tunnel Coolers.

C.

• An Operability Test demonstrates that Service Water can supply 300,000 Gallons of water from the Cooling Tower.

• A Functional Test demonstrates that the Diesel Generator can carry 4 MWe load.

D.

• An Operability Test demonstrates that the Clarified Water System can provide 100,000 Gallons of water to the Demin Water header.

• A Functional Test demonstrates that the HPCI System can inject within 10 seconds of an initiation signal.

LOC-23 NRC Exam Rev 3

K&A# 2.2.21 Importance Rating 4.1 QUESTION 98 SRO Tier 3 K&A Statement: Knowledge of pre- and post-maintenance operability requirements.

Justification:

A. Incorrect, Service water flow is not a safety function, RCIC flow is a safety function.

Candidates may select this if they incorrectly recall system safety functions.

B. Correct, per NDAP-QA-0482 ... "OPERABILITY TEST - Activities or tests which ensure that the structures, systems, components affected by the maintenance or modification activities are capable of providing the safety functions specified in the Current Licensing Basis. FUNCTIONAL TEST - Activities or tests which ensure that structures, systems, components meet design performance requirements. Generally applies to non-Tech Spec equipment, where operability testing is not applicable. "

The demonstration that RCIC can deliver its design flow is an activity intended to verify safety function capability. The Service Water flow to Main Steam tunnel coolers is NOT designed to verify a safety function.

C. Incorrect, Service water flow is not a safety function, DG loading is a safety function.

Candidates may select this if they incorrectly recall system safety functions.

D. Incorrect, Clarified water flow is not a safety function, HPCI injection response time is a safety function. Candidates may select this if they incorrectly recall system safety functions.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SRO candidates must recall procedural rules for post maintenance operability testing. Ensuring that PMT is sufficient to determine safety system operability is strictly an SRO function.

References:

NDAP-QA-0482 rev 4 Reference Required None Learning Objective: 15018 Question source: SSES OPS_INITIAL_L1CENSE Bank #AD044/15018 002 Question History: Bank Cognitive level: Memory/Fundamental knowledge: X Comprehension!Analysis:

10CFR55 43(b)2 Comments: Created!Modified by: Bank Reviewed by: T. Ebert, L. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 99 SROONLY Unit 1 is operating at full power when it is determined that the following Standby Liquid Control (SBlC) tank chemistry conditions currently exist:

• Tank temperature is 67°F

• Sodium Pentaborate concentration is 9.0 weight-%

• Tank volume is 1400 gallons Given Tech Spec 3.1.7, determine which one of the following is correct:

A. The pH of the suppression pool may exceed 7.0 following an ATWS concurrent with a lOCA because sodium pentaborate concentration is too low.

B. Re-evolution of iodine in the suppression pool may occur following a Design Basis lOCA because sodium pentaborate concentration is too high.

C. Cold shutdown will be achieved during ATWS conditions because sodium pentaborate concentration meets requirements for the given tank volume and temperature.

D. Cold shutdown may NOT be achieved during an ATWS because the sodium pentaborate concentration is too low.

LOC-23 NRC Exam Rev 3

K&A# 2.1.34 Importance Rating 3.5 QUESTION 99 SROTier3 K&A Statement: Knowledge of primary and secondary plant chemistry limits.

Justification:

A. Incorrect, suppression pool pH is a concern only during DBA lOCA. In that case, SBlC is used to keep pH above 7.0, vice below 7.0 as this distractor implies.

Candidates may select this if they incorrectly apply the SP pH basis for the boron concentration tech spec.

B. Incorrect, although this is a valid result of the current conditions, tank volume is currently too lOW for the given concentration of solution. Candidates may select this if they incorrectly plot the given variables in the TS graph.

C. Incorrect, tank boron concentration is not acceptable for the given conditions.

Candidates may select this if they incorrectly plot the given variables in the TS graph.

D. Correct, sodium pentaborate solution is lower than required. With inadequate solution, the SBLC system may not perform its safety.

KIA Match & SRO Only Justification:

This question matches the stated KIA since SROs must correctly apply the standby liquid control system chemistry requirements, and correctly recall the TS basis for those requirements.

References:

TS 3.1 .7 rev 3, TSB 3.1.7 rev 3 Reference Required TS 3.1.7 learning Objective: 10099 Question source: MODIFIED SSES OPS_INITIAl_L1CENSE Bank

1. TM0053/10099 003 Question History: MODIFIED Bank Cognitive level: Memory/Fundamental knowledge:

Comprehension/Analysis: X 10CFR55 43(b)5 Comments: Modified by: T. North, 7/5/10 Reviewed by: T. Ebert, L.. Casperson 11-12-10 LOC-23 NRC Exam Rev 3

QUESTION 100 SROONLY Unit 1 is operating at 2% power.

Maintenance personnel have entered the primary containment to perform emergent repairs on elevation 738'.

Due to anticipated Xenon burnout reactor power begins to rise slowly.

Given NDAP-QA-0309, "Primary Containment Access & Control", determine what action (if any) the crew should take:

A. Place the mode switch to SHUTDOVVN.

B. No action is necessary until reactor power approaches 10%.

C. Direct all personnel to immediately exit the drywell.

D. Insert Control Rods to maintain reactor power S 3%.

LOC-23 NRC Exam Rev 3

K&A# 2.3.13 Importance Rating 3.7 QUESTION 100 SROTier3 K&A Statement: Radiation Control: Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc Justification:

A. Incorrect, since the power change is anticipated, placing the mode switch to shutdown is not required.

B. Incorrect, power must be maintained ::>~~% with personnel on elev. 738'.

C. Incorrect, personnel are permitted to remain in the drywell provided power is maintained ::>3%.

D. Correct, per NDAP-QA-0309, reactor power is restricted to :s 3% for access to drywell elevation 738'. Control rod insertion is permitted with personnel inside containment and power below 3%.

KIA Match & SAO Only Justification:

This question matches the stated KIA since SRO candidates must evaluate plant conditions to determine the correct action in accordance with the primary containment access procedure.

References:

NDAP-QA-0309, rev 26 section 6.5 Reference Required none Learning Objective: 15314 Question source: New Question History: New Cognitive level: Memory/Fundamental knowledge: X Comprehension/Analysis:

10CFR55 43(b)4 Comments: Created by: T. North, 11/16/10 Reviewed by:

LOC*23 NRC Exam Rev 3