ML17023A123

2016 Monticello Nuclear Generating Plant Initial Licensed Operator Written Examinations

ML17023A123
Person / Time
Site:	Monticello
Issue date:	01/20/2017
From:	NRC/RGN-III
To:	Nuclear Management Co
Shared Package
ML15274A425	List: ML17023A109 ML17023A113 ML17023A118 ML17023A123 ML17121A169
References
Download: ML17023A123 (301)

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FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 1 Points: 1.0 The reactor was operating at rated conditions when the 11 Reactor Recirc pump tripped. Three minutes later, indicated Total Core Flow on Panel C-04 recorder is 38 Mlbm/hr.

Actual core flow is assumed to be A. higher than indicated due to reverse flow in the idle loop.

B. lower than indicated due to reverse flow in the idle loop.

C. lower than indicated due to lower reactor power resulting in a lower core P.

D. the same as indicated due to a subtraction circuit in the Total Core Flow summing logic.

Monticello NRC Written Exam Page: 1 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 1 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A. Incorrect, this is the opposite effect B. Correct, Total core flow is a sum of recirc loop flows. Since reverse flow may occur in the idle loop during high power/recirc pump trip conditions, this reverse flow is added to the forward flow in the running loop and makes indicated flow appear higher than actual core flow.

C. Incorrect, Core flow is equally affected by core P.

D. Incorrect, this logic is not installed at MNGP QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 295001 Partial or Complete Loss of Forced Core Flow Circulation AA2.03 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Actual core flow Importance Rating: 3.3 10 CFR Part 55: 55.41.10 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine the effect of a recirc pump trip on actual core flow indication SRO Justification: N/A Technical

References:

Ops Man C.4-B.05.01.02.A Proposed references None to be provided:

Learning Objective: MT-ILT-AOP-003L Monticello NRC Written Exam Page: 2 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 1 Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank _

Bank X Question History: MNGP 2015 ILT Audit exam Comments:

Monticello NRC Written Exam Page: 3 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 2 Points: 1.0 The plant was at rated conditions when an event occurred. Given the following:

• A Station Blackout and LOCA have occurred

• 13 Diesel Generator is supplying LC-107 and LC-108 (Back feed NOT available)

• C.5-1200 (PRIMARY CONTAINMENT CONTROL) has been entered Which of the following actions from C.5-1200 can be accomplished (assume manual operation of MOTOR OPERATED valves is performed when required)?

A. Lower Torus water level IAW C.5-3402 (DRAINING TORUS WATER TO RADWASTE).

B. Establish H 2 /O 2 Drywell sampling IAW C.5-3501 (H 2 /O 2 ANALYZER OPERATION).

C. Use Fire Water System to initiate Drywell sprays IAW C.5-3502 (CONTAINMENT SPRAY).

D. Reduce Drywell air temperature IAW C.5-3503 (DEFEAT DRYWELL COOLER TRIPS).

Monticello NRC Written Exam Page: 4 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 2 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A. Incorrect: a prerequisite to perform C.5-3402, Draining Torus Water to Radwaste, is having RHR available for service; however, with a SBO, and only the non-essential 13 Diesel Generator available and only powering to LC-107 and LC-108, RHR is NOT available.

B. Incorrect: H2/O2 Analyzers are NOT available as they are powered from MCC-133A and MCC-143A (LC-103 and LC-104) via Power Panels P-73A and P-73B.

C. Correct: per EOP C.5-3502, IF a RHR pump is not available, THEN several options (PARTs) to spray containment are available, but, given the conditions, PART H of C.5-3502, USE FIRE WATER SYSTEM FOR CONTAINMENT SPRAY is the only viable option.

All MOVs are in the Reactor Building and are accessible for manual operation.

D. Incorrect: DW Cooling Fans are NOT available as they are powered from MC-133A and MCC-143A (LC-103 and LC-104)

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (295003 Partial or Complete Loss of AC / 6) Generic 2.4.6 Knowledge of EOP mitigation strategies. (CFR: 41.10 / 43.5 / 45.13)

Importance Rating: 3.7 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the EOP mitigating strategies available when the plant experiences a loss of AC SRO Justification: N/A Technical

References:

Ops Man C.4-B.09.02.A, EOP C.5-1200 and -3502, M-8107L-023 (RHR System) Rev. 20, M-8107L-106 (E-Manual) Rev. 6, M-8107L-040 (480 Volt System) Rev. 14, M-8107L-089 (No. 13 Diesel Generator) Rev. 8 Monticello NRC Written Exam Page: 5 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 2 Proposed references to be provided: None Learning Objective: 5.k of M-8107L-023 (RHR System) Rev. 20 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: Question 1952 from the Exam BANK Comments:

Monticello NRC Written Exam Page: 6 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 3 Points: 1.0 On a loss of Division 1 125 VDC, which of the following will be non-functional?

A. Main Generator lockout relay trips B. Main Turbine vacuum trip C. 12 Recirc MG Set protective trip circuits D. HPCI remote turbine trip Monticello NRC Written Exam Page: 7 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 3 ANSWER EXPLANATION Answer: A DISTRACTOR ANALYSIS A. Correct.

B. Incorrect, Turbine controls vacuum trip is non-functional on a loss of 125 VDC Bus B C. Incorrect, Recirculation MG Set No. 12 protective trip circuits are non-functional on a loss of 125 VDC Bus B D. HPCI remote turbine trip function is non-functional on a loss of 125 VDC Bus B QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 295004 Partial or Total Loss of DC Pwr / 6.

AK1.05 Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER : Loss of breaker protection Importance Rating: 3.3 10 CFR Part 55: 55.41.8 to 41.10 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine that the main generator lockout relay trips are non-functional. The lockout relay trips are part of the breaker protection relaying for the 8N7 and 8N8 breakers that provide connection for the main transformer to the Xcel 345kV system.

SRO Justification: N/A Technical

References:

Abnormal Procedures: LOSS OF A 125 V DC BUS Ops Man C.4-B.09.10.A Revision 14.

M-81 07L-036 (345 kV, 230 kV & 115 kV Substation) Rev. 20 M-8107L-039 (4.16 kV And 13.8 kV Station Auxiliary Systems) Rev. 21 Proposed references to be provided: None Monticello NRC Written Exam Page: 8 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 3 Learning Objective: M-81 07L-036 (345 kV, 230 kV & 115 kV Substation) Rev. 20

5. Describe the functional and component relationships between the 345 kV, 230 kV, and 115 kV Substation and the following: a. 125 Vdc Distribution System Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History: New Comments:

Monticello NRC Written Exam Page: 9 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 4 Points: 1.0 A plant shutdown is in progress.

• Reactor power is 25%

• 5-B-37 (TURBINE PRESS GEN REJECTION BYPASS) is in alarm

• 5A-K9A (Pressure Bypass Load Reject & Stop Valve relay) is energized

• 5A-K9B (Pressure Bypass Load Reject & Stop Valve relay) is de-energized

• 5A-K9C (Pressure Bypass Load Reject & Stop Valve relay) is de-energized

• 5A-K9D (Pressure Bypass Load Reject & Stop Valve relay) is de-energized Given the above conditions and prints NX-7834-67-9(10), if the Main Turbine is tripped, what would be the effect on RPS?

A. No RPS actuation B. Half Scram on RPS Channel A C. Half Scram on RPS Channel B D. Full Scram Monticello NRC Written Exam Page: 10 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 4 ANSWER EXPLANATION Answer: D DISTRACTOR ANALYSIS A, B, and C are incorrect but plausible interpretations of the effects on RPS.

D is correct: ARP 5-B-37 directs the operator to check that the relays in the control room are energized. Memorization of ARP steps is not required and prints are provided to justify operational validity. Current status of the Pressure Bypass Load Reject & Stop Valve relays are indicative of -

only one bypass relay is currently energized (5A-K9A on Panel C-15). In accordance with B.05.06-02, to provide a bypass for the stop valve and control valve closures while at low-power, four non-indicating pressure switches are used to monitor the first stage turbine pressure and will provide a bypass for the scram trips when first stage pressure is below 25% of its rated full power pressure (~118 psig). Correspondingly, according to ARP 5-B-37 and Form 2204, alarm 5-B-37 is designed to alarm below 25% reactor power AND the alarm is actuated if ANY turbine first stage pressure switch reaches pressure of ~95 psig. In this case, since 3 relays are still de-energized (RPS Channels A2, B1, and B2), the RPS trip from the turbine trip would NOT be bypassed and a scram would result.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (295005 Main Turbine Generator Trip / 3) AK2.01 Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: (CFR: 41.7 / 45.8)

RPS Importance Rating: 3.8 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the association between the main turbine generator trip and RPS SRO Justification: N/A Monticello NRC Written Exam Page: 11 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 4 Technical

References:

Ops Man B.05.06-02 and -06, ARP 5-B-37, Form 2204, NX-7834-67-9(10),

M-8107L-072 (Reactor Protection System) Rev. 16, M-8107L-013 (Main Turbine) Rev. 16 Proposed references to be provided: NX-7834-67-9(10)

Learning Objective: 4.c of M-8107L-072 (Reactor Protection System) Rev. 16 Cognitive Level: Higher X Lower Question Source: New _

Modified Bank _

Bank X Question History: Question 1245 from the MNGP Exam BANK Comments: Revised proposed question to provide reference for operational validity. A control room operator would not be expected to know from memory that specific relays are energized/deenergized when an alarm is received.

Monticello NRC Written Exam Page: 12 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 5 Points: 1.0 Complete the following statement describing the immediate reactor water level response to a manual scram from rated conditions AND the reason?

RPV water level will...

A. lower due to the Recirc pump runback.

B. rise due to delay in Bypass Valves opening.

C. lower due to collapsing voids in the core region.

D. rise due to rapid closure of the Turbine Control Valves.

Monticello NRC Written Exam Page: 13 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A. Incorrect, Recirc pumps should runback to 30% but this tends to raise reactor level.

B. Incorrect, Bypass valves opening will raise water level but since reactor pressure also lowers on a scram these valves would not be expected to open immediately.

C. Correct D. The turbine control valves will close but this would cause level to lower.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 295006 SCRAM / 1 AK3.01 Knowledge of the reasons for the following responses as they apply to SCRAM : Reactor water level response...

Importance Rating: 3.8 10 CFR Part 55: 55.41.5 10 CFR 55.43.b  : N/A K/A Match: On a manual scram from rated conditions, the sudden decrease in thermal power causes the voids in the core to collapse. This causes the indicated level to lower.

SRO Justification: N/A Technical

References:

C.4-A Bases Proposed references None to be provided:

Learning Objective: MT-ILT-AOP-001L Monticello NRC Written Exam Page: 14 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 5 Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank _

Bank X Question History: NRC 2010 exam MNGP Comments:

Monticello NRC Written Exam Page: 15 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 6 Points: 1.0 The plant was at rated conditions when an event occurred that required Control Room evacuation.

• An emergency depressurization was performed to restore RPV water level.

• RPV pressure is 70 psig and lowering.

• Alternate Shutdown System (ASDS) Fuel Zone RPV water level indicator reading is as shown:

Based on the given conditions, complete the following statement concerning RPV water level?

Actual RPV water level is ______ than indicated because this instrument is calibrated at ____ .

A. lower, 0 psig B. lower , 1000 psig C. higher, 0 psig D. higher, 1000 psig Monticello NRC Written Exam Page: 16 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 6 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A, C, and D are incorrect but plausible misconceptions of water level indications during off-normal conditions. According to Ops Man B.01.01-01, Level transmitter LT-2-3-61, which is located on C-55 and used for flooding the vessel during shutdown from outside the Control Room, is the only level transmitter that is calibrated for COLD conditions (0 psig).

B is correct: The examinee must understand that the Fuel Zone level indicator on the ASDS panel C-292, LI-4108, is calibrated for 1000 psig (HOT). With RPV pressure at 70 psig, cold conditions are established. The colder water requires less inventory to provide the same level indication.

Therefore, actual water level is lower. As plotted on Figure 29 of B.01.01-06 actual water level is approximately -65 inches. The fuel zone level instrumentation will NOT be reliable during the depressurization phase of a LOCA.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (295016 Control Room Abandonment / 7) AA1.06 Ability to operate and/or monitor the following as they apply to CONTROL ROOM ABANDONMENT: (CFR: 41.7 /

45.6) Reactor water level Importance Rating: 4.0 10 CFR Part 55: 41.7 10 CFR 55.43.b  : N/A K/A Match: K/A is matched by testing the applicants ability to accurately monitor the reactor water level when outside the control room.

SRO Justification: N/A Technical

References:

Ops Man B.01.01-01, -03, and -06, C.4-C, and M-8107L-083 (Alternate Shutdown System) Rev. 9 Proposed references to be provided: None Monticello NRC Written Exam Page: 17 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 6 Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: Question 1870 from the Exam BANK Comments: The question stem has been slightly modified from the original BANK question in order to improve presentation.

Monticello NRC Written Exam Page: 18 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 7 Points: 1.0 The plant is at rated conditions with 11 RBCCW pump in service.

Electrical power is momentarily lost to 12 RBCCW Pump resulting in the following conditions:

If 6-B-32 (RBCCW LOW DISCH PRESS) is received and the condition of 11 RBCCW Pump remains as shown above, which of the following is correct?

A. 11 RBCCW pump may have a sheared shaft 12 RBCCW pump WILL auto start B. 11 RBCCW pump may have a sheared shaft 12 RBCCW pump will NOT auto start C. 11 RBCCW pump discharge valve automatically closed 12 RBCCW pump WILL auto start D. 11 RBCCW pump discharge valve automatically closed 12 RBCCW pump will NOT auto start Monticello NRC Written Exam Page: 19 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 7 ANSWER EXPLANATION Answer B DISTRACTOR ANALYSIS An 11 RBCCW pump shaft shear will result in a low discharge header condition alarm at 30 psig and the pump motor could be still running as indicated by the red indicating lamp. 12 RBCCW pump is shown without the standby setup feature enabled (white light off) so it will fail to start.

A. (1) Correct (2) Incorrect - The 12 RBCCW Pump will not auto start.

B. (1) Correct (2) Correct C. (1) Incorrect - The discharge valve is a manual valve and has no automatic features.

(2) Incorrect - The 12 RBCCW Pump will not auto start.

D. (1) Incorrect - The discharge valve is a manual valve and has no automatic features.

(2) Correct QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 295018 Partial or Total Loss of CCW / 8 AA2.03 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER : Cause for partial or complete loss...

Importance Rating: 3.2 10 CFR Part 55: 55.41.10 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine that the setup feature of the standby RBCCW pump is lost when the white light is off. The applicant is asked the low discharge header pressure that would result in a standby pump start if in setup and identify the status of the standby pump.

Monticello NRC Written Exam Page: 20 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 7 SRO Justification: N/A Technical

References:

B.02.05-02 C-4-B.02.05.A ARP 6-B-32 Proposed references None to be provided:

Learning Objective: M-8107L-026 LO 7 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Modified from 2015 MNGP ILT NRC question 7 Comments:

Monticello NRC Written Exam Page: 21 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 8 Points: 1.0 The plant is at rated conditions when an event occurs resulting in the common Service/Instrument air header pressure slowly lowering.

Assuming the common air header pressure continues to lower; which of the following alarms would be the FIRST to be received that ALSO requires entry into an AOP (Abnormal Operating Procedure)?

A. ROD DRIFT (5-A-27)

B. INST AIR HEADER LOW PRESS (6-B-34)

C. SERVICE AIR HEADER LOW PRESS (6-B-35)

D. SCRAM PILOT HEADER HI/LO PRESS (5-B-22)

Monticello NRC Written Exam Page: 22 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 8 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A. Incorrect: at <60 psig, the control rods will start to drift and the reactor must be manually scrammed; but AOP shall be entered prior to.

B. Correct: when Air Header Pressure reaches 85 psig, annunciator 6-B-34 (INST AIR HEADER LOW PRESS) will alarm. ARP 6-B-34 directs AOP C.4-B.08.04.01 entry. Per AOP C.4-B.08.04.01, Table 1, if the IA pressure is reduced to 85 psig this is the first time operator AOP action will be required.

C. Incorrect: at 82 psig, Annunciator 6-B-35 (SERVICE AIR HEADER LOW PRESS) alarms and CV-1474, Serv Air Isol CV valve, closes; AOP entry is not directly required.

D. Incorrect: Annunciator 5-B-22 (SCRAM PILOT HEADER HI/LO PRESS) alarms at 60 psig. AOP entry would not be required until control rods start to drift.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (295019 Partial or Total Loss of Inst. Air / 8) 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures. (CFR: 41.10 /

43.2 / 45.6)

Importance Rating: 4.5 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants ability to recognize Instrument Air pressure has reduced to an AOP entry level condition SRO Justification: N/A Technical

References:

AOP C.4-B.08.04.01.A, ARPs 6-B-34, 6-B-35 and 5-B-22, and M-8107L-024 (Instrument and Service Air Alternate Nitrogen and Instrument Nitrogen Systems) Rev. 24 Proposed references to be provided: None Monticello NRC Written Exam Page: 23 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 8 Learning Objective: 7 of M-8107L-024 (Instrument and Service Air Alternate Nitrogen and Instrument Nitrogen Systems) Rev. 24 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History: NEW - Question 78 from 2013 MNGP ILT NRC exam used as a starting point but converted from SRO to RO question Comments:

Monticello NRC Written Exam Page: 24 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 9 Points: 1.0 A plant startup is in progress with the following conditions:

• RPV water level is being maintained +30 to +35 inches

• RHR SDC is in service If a LONOP subsequently occurs, complete the following statements?

(1) RHR SDC will _________.

(2) Maintaining RPV water level in the above band _________ natural circulation within the core.

A. (1) remain in service (2) ensures B. (1) remain in service (2) does NOT ensure C. (1) be lost (2) ensures D. (1) be lost (2) does NOT ensure Monticello NRC Written Exam Page: 25 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 9 ANSWER EXPLANATION Answer: D DISTRACTOR ANALYSIS A. (1) Incorrect, A Group 2 PCT isolation logic will occur due to the loss of normal offsite power, this results in a loss of SDC (2) Incorrect, Maintaining a reactor water level of +55 or greater ensures that circulation can occur between the inside and outside of the core shroud B. (1) Incorrect, Maintaining a reactor water level of +55 or greater ensures that circulation can occur between the inside and outside of the core shroud.

(2) Correct C. (1) Correct (2) Incorrect, Maintaining a reactor water level of +55 or greater ensures that circulation can occur between the inside and outside of the core shroud D. Correct QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 295021 Loss of Shutdown Cooling / 4.

AK1.04 Knowledge of the operational implications of the following concepts as they apply to LOSS OF SHUTDOWN COOLING : Natural circulation Importance Rating: 3.6 10 CFR Part 55: 55.41.8 to 41.10 10 CFR 55.43.b: N/A K/A Match: Plant startup requires a RPV level band of +30 to +35, SDC or a recirc pump is required to be in service during this Condition, a loss of normal offsite power will result in a loss of both and the RPV level will not support natural circulation SRO Justification: N/A Monticello NRC Written Exam Page: 26 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 9 Technical

References:

Shutdown Procedure, Ops Man C.3 Revision 79; LOSS OF NORMAL SHUTDOWN COOLING, Ops Man C.4-B.03.04.A; 2167 Plant Startup Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History: New Comments:

Monticello NRC Written Exam Page: 27 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 10 Points: 1.0 The plant is operating in MODE 1 with fuel bundles being shuffled in the Spent Fuel Pool. An event occurs resulting in the following conditions:

• 4-A-1 (REFUELING FL AREA HI RADIATION) is in alarm

• ARM A-1 (1027 RB NORTH EAST LOW) is in alarm

• ARM A-2 (1027 RB NORTH HIGH) is in alarm

• ARM A-3 (1027 RB WEST STAIRWAY) is in alarm

• Refuel Bridge radiation monitor is in alarm

• RM-17-453A (Fuel Pool Monitor Channel A) is 30 mr/hr and slowly rising

• RM-17-453B (Fuel Pool Monitor Channel B) is 25 mr/hr and slowly rising Which one of the following is the correct response to these conditions?

A. Evacuate the Refuel Floor ONLY.

B. Verify automatic initiation of SBGT ONLY.

C. Evacuate the Reactor Building AND the Refuel Floor.

D. Verify automatic initiation of SBGT and CRV-EFT (High Radiation Mode).

Monticello NRC Written Exam Page: 28 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 10 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A. Incorrect: The RB should also be evacuated in accordance with Ops Manual D.2-05. If only the Refuel Bridge radiation alarm sounds, RB evacuation is not required.

B. Incorrect: SBGT would not be initiated automatically until RM-17-453A or B reached 50 mr/hr.

C. Correct: When 4-A-1 alarms due to ARM A-1, A-2, and/or A-3 on C-11, the ARM indicators should be checked and it shall be determined the affected area and radiation level.

According to Ops Man B.05.12-03, ONLY ARM A-1 and A-3 also have a local range indication at the sensor with an upscale trip light and horn to indicate an alarm condition.

Since BOTH RB and Refuel Bridge radiation alarms are sounded, according to Ops Man D.2-05 and ARP 4-A-1, all personnel in the RB and Refuel Floor Area shall evacuate.

D. Incorrect: See B above.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (295023 Refueling Acc / 8) AK2.03 Knowledge of the interrelations between REFUELING ACCIDENTS and the following: (CFR: 41.7 / 45.8) Radiation monitoring equipment Importance Rating: 3.4 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the association between refueling accidents and radiation monitoring equipment SRO Justification: N/A Technical

References:

Ops Man B.05.12-01 and -03, ARP 4-A-1, Ops Man D.2-05, M-8107L-019 (Fuel Handling) Rev. 14, M-8107L-077 (Process Radiation Monitoring And Area Radiation Monitors) Rev. 15 Proposed references to be provided: None Monticello NRC Written Exam Page: 29 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 10 Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Question 10 from the 2010 Audit Exam Comments: Minor editorial changes were made and distracter D reworded slightly.

Monticello NRC Written Exam Page: 30 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 11 Points: 1.0 The plant was at rated conditions when a transient occurred.

• Torus water temperature is 107°F and rising

• Drywell (DW) pressure is 40 psig and rising

• DW temperature is 289°F and rising

• DW Sprays are unavailable The CRS has directed DW Venting IAW C.5-3505 (VENTING PRIMARY CONTAINMENT).

What is the reason for venting the DW with the conditions above?

A. To raise the margin to the RHR NPSH Limit B. To raise the margin to the Heat Capacity Limit C. To control pressure below the Drywell Pressure Limit D. To lower DW temperature below the ADS EQ Temperature Limit Monticello NRC Written Exam Page: 31 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 11 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS The parameters and trends in the stem replicate a situation where a significant LOCA occurred in the DW. Assuming all RPV injection sources are available and RPV level is controlled it is feasible to assume the crew would anticipate a blowdown to the main condenser to limit the water temperature rise in the torus. If containment pressure continues to rise following the blowdown, the containment is vented to control pressure below the DW Pressure Limit.

A. Incorrect; without references it can be fundamentally determined that NPSH would not be a concern with a high drywell pressure and venting would lower the margin.

B. Incorrect; with Torus water temperature still below the TS required scram limit it can be fundamentally determined that the Heat Capacity limit will not be exceeded.

C. CORRECT D. Incorrect; the DW temperature ADS EQ temperature limit is 338° which is not being exceeded QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 295024 High Drywell Pressure / 5.

EK3.07 Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: Drywell venting Importance Rating: 3.5 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A Monticello NRC Written Exam Page: 32 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 11 K/A Match: The K/A is matched as the applicant must evaluate primary containment conditions against the design basis and evaluate the mitigate strategy, such that venting the drywell is permitted to be performed prior to spraying the drywell to reduce total dose.

SRO Justification: N/A Technical

References:

USAR-05.02 C.5.1-1200 C.5-1200 Proposed references to be provided: None Learning Objective: M-8107L-044 LO 9,11 Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History: New Comments:

Monticello NRC Written Exam Page: 33 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 12 Points: 1.0 The plant was at rated conditions when a loss of feedwater resulted in a reactor scram and RCIC/HPCI auto-initiation on LO-LO RPV Level.

• MSIVs are closed and RPV pressure is rising

• HPCI failed to shutdown on High RPV water level

• RPV water level has reached the main steam lines

• The CRS directs manual RPV pressure control with SRV(s)

• The SRV F handswitch is taken to OPEN, but its amber light does NOT illuminate Which of the following actions should be taken next?

A. Place SRV F hand-switch in CLOSE and then any other SRV hand-switch to OPEN.

B. Cycle SRV F hand-switch to CLOSE and immediately back to OPEN.

C. Monitor RPV pressure and SRV tailpipe temperature to determine if SRV F is open.

D. Place SRV hand-switches in OPEN following the preferred sequence until at least one SRV amber light illuminates.

Monticello NRC Written Exam Page: 34 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 12 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A. Incorrect: according to B.03.03-05.H, SRV B is NOT the next preferred sequence after SRV F - SRV C is B. Incorrect: Procedurally incorrect: According to B.03.03-05.H, Once a relief valve has been closed, it should NOT be reopened for at least 10 seconds (in the case of SRV E, G, and H, reopening is blocked for 10 seconds after alarm clears) or discharge piping and containment damage may result C. Correct: In this condition, a reactor scram has occurred, the MSIVs are closed and Reactor pressure is increasing - and the CRS directs use of the SRV(s) as an alternate pressure control system (per C.5-1100, RPV Control). SRV open indication (amber light on) may NOT be received or could be delayed significantly during SRV actuation if water is present in the SRV or associated steam line versus an SRV actuation with steam only. SRV tailpipe temperature changes and a decrease in Reactor pressure are both means to monitor SRVs and they should be reviewed prior to opening the next SRV in the sequence (a sequence is identified in Ops Man B.03.03-05.H that shall be employed to spread the heat around the Torus and will open the SRVs with a lower number of life cycles first).

D. Incorrect: Procedurally not the preferred approach; it may take some time for the amber light to illuminate and there are indications that should be reviewed before another SRV is opened to minimize pressure transients.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (295025 High Reactor Pressure / 3) EA1.03 Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE: (CFR: 41.7 / 45.6)

Safety/relief valves: Plant-Specific Importance Rating: 4.4 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants ability to operate the SRV(s) according to procedure when a high reactor pressure condition exists AND accordingly be able to monitor the SRV(s) status. The applicant shall understand what indications actually represent - the amber lights do NOT indicate actual SRV position.

Monticello NRC Written Exam Page: 35 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 12 SRO Justification: N/A Technical

References:

Ops Man B.03.03-01 and -05, EOPs C.5-1100 and -3302, ARP 5-A-46, M-8107L-025 (Reactor Pressure Relief) Rev. 23 Proposed references to be provided: None Learning Objective: 7 of M-8107L-025 (Reactor Pressure Relief) Rev. 23 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: Question 549 of Exam BANK - with the question stem and distractors modified in an attempt to strengthen question and more-appropriately match K/A.

Comments:

Monticello NRC Written Exam Page: 36 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 13 Points: 1.0 The plant was at rated conditions when a SBO and ELAP event occurred. Given the following:

• The Heat Capacity Limit has been exceeded

• RPV water level is being maintained with HPCI & RCIC

• The CRS has entered C.5-2002 (EMERGENCY DEPRESSURIZATION)

During the depressurization, the CRS directs a pressure band of 150-200 psig.

Is the directed pressure band correct? Why or why not?

A. CORRECT; >150 psig is required to maintain core cooling.

B. CORRECT; >150 psig is required to mitigate primary containment challenges.

C. INCORRECT; full depressurization is required to maintain core cooling.

D. INCORRECT; full depressurization is required to mitigate primary containment challenges.

Monticello NRC Written Exam Page: 37 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 13 ANSWER EXPLANATION Answer: A DISTRACTOR ANALYSIS A. Correct: Under circumstances requiring ED, it is generally desirable to full depressurize the RPV. However, while RPV pressure reductions will tend to increase flow from motor-driven pumps, full depressurization will result in loss of steam driven injection sources. RPV pressure reduction must be coordinated with core cooling strategies. Full depressurization is only appropriate if adequate core cooling will not be sacrificed as a result. In the ELAP condition HPCI and RCIC will be the only injection sources so RPV pressure is to be maintained > 150 psig.

B. Incorrect: The pressure band is not for containment concerns.

C. Incorrect: Full depressurization not required.

D. Incorrect: The pressure band is not for containment concerns and full depressurization not required.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 295026 Suppression Pool High Water Temp.

EA2.03 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor pressure Importance Rating: 3.9 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: The applicant must properly determine the required pressure strategy for high torus water temperature conditions requiring an ED.

Monticello NRC Written Exam Page: 38 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 13 SRO Justification: N/A Technical

References:

C.5.1-2002, C.5-2002, C.5-4000 Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History: New Comments:

Monticello NRC Written Exam Page: 39 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 14 Points: 1.0 The plant was at rated conditions with Drywell Cooling Fan 1 in STANDBY when a Loss of ALL Off-site Power occurred.

• RPV Water Level is +15 inches, lowest level reached was -35 inches

• Control Rod Drive Area temperature is 120°F and stable

• Drywell Temperature is 135°F and slowly rising

• Drywell Pressure 1.2 psig and slowly rising What actions, if any, are required to place ALL AVAILABLE Drywell Cooling in service?

A. No actions are required; all available Drywell Cooling is already in service.

B. Open ALL four DW Cooling Fan Discharge Dampers.

C. Place ALL DW Cooling Fan control switches to OFF; Verify Fan Inlet Damper control switches in AUTO; Place ALL DW Cooling Fan control switches to ON; Open associated DW Cooling Fan Discharge Dampers.

D. Verify DW Cooling Fans 2, 3, & 4 restarted and Fan Inlet Dampers are open; Open DW Cooling Fans 2, 3, & 4 Discharge Dampers; Start DW Cooling Fan 1 and verify Inlet Damper opens; Open DW Cooling Fan 1 Discharge Damper.

Monticello NRC Written Exam Page: 40 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 14 ANSWER EXPLANATION Answer: D DISTRACTOR ANALYSIS A. Incorrect; DW Cooling Fans will restart following restoration of AC power by the EDGs, but fan discharge dampers have to be manually opened.

B. Incorrect; DW Cooling Fan 1 was in Standby prior to the power loss and therefore does not restart.

C. Incorrect; DW Cooling Fans 2, 3, and 4 will already be running therefore stopping the fans, then restarting is unnecessary. DW Cooling Fan 1 is in Standby and starting the fan does not require placing the control switch to OFF and then to ON.

D. Correct; DW Cooling Fans 2, 3, and 4 will be running, but discharge dampers must be opened. DW Cooling Fan 1 must be started and then its discharge damper opened.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (295028 High Drywell Temperature / 5) 2.2.44 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. (CFR: 41.5 / 43.5 / 45.12)

Importance Rating: 4.2 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: Applicant must evaluate given plant conditions to determine DAC system status and then determine the actions necessary to ensure all available Drywell Cooling is in service.

Monticello NRC Written Exam Page: 41 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 14 SRO Justification: N/A Technical

References:

B.08.16-01/02 Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 42 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 15 Points: 1.0 The plant was preparing to shutdown due to low torus water level from a leak that has been isolated. While still at rated conditions a reactor scram occurred due to an inadvertent Group I isolation. Control rods FAILED to insert and the following plant conditions now exist:

• Reactor power is 10% and slowly lowering due to SBLC injection

• Torus water temperature is 160°F

• ALL available torus cooling is in service

• Torus water level is -2 ft.

• Drywell pressure is 3.5 psig Which of the following states the correct adjustment, if any, to torus cooling necessary to maintain maximum available torus cooling and stay within pump operating limits?

A. No adjustment required B. Reduce flow in each RHR loop to 4000 gpm C. Reduce flow in each RHR loop to 6000 gpm D. Reduce flow in each RHR loop to 7500 gpm Monticello NRC Written Exam Page: 43 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 15 ANSWER EXPLANATION Answer A DISTRACTOR ANALYSIS A is correct as torus overpressure is calculated to be approximately 4.5 psig. RHR flow should be operated at the maximum flow rate possible to reduce torus temperature. Interpolation is allowed for Detail F; therefore 8000 gpm loop flow wouldnt exceed NPSH limits until approximately 180°F torus water temperature.

A. Correct B. Incorrect; If 0 psig curve is used, it is under the 2 pumps curve but does not maximize available Torus Cooling.

C. Incorrect, plausible if 0 psig curve is used.

D. Incorrect, this it is too much flow, but is plausible if 0 psig curve is used since flow corresponds to limit for 1 pump QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 295030 Low Suppression Pool Wtr Lvl EK1.02 Knowledge of the operational implications of the following concepts as they apply to LOW SUPPRESSION POOL WATER LEVEL: Pump NPSH Importance Rating: 3.5 10 CFR Part 55: 55.41.8 to 10 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine the required action to maintain the RHR system operation within NPSH requirements.

SRO Justification: N/A Technical

References:

C.5-GM-07, C.5-1200 Figure F, C.5.1-1200, C.5-1200 Monticello NRC Written Exam Page: 44 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 15 Proposed references to be provided: C.5-1200 Figure F (inserted with question)

Learning Objective: M-8107L-044, LO 8 &10 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Modified NRC exam 2015 Q 13, changes to stem to address low torus water level and 0 psi overpressure.

Comments:

Monticello NRC Written Exam Page: 45 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 16 Points: 1.0 Which of the following completes the statement below?

The Reactor Protection System reactor low water level signal is received by __(1)__ level transmitters AND will send a SCRAM input __(2)__ time delay if the +9 inch setpoint is indicated.

(1) (2)

A. 4 independent WITHOUT a B. 4 independent WITH a 7.3 second C. 2 independent and 1 Safeguard WITHOUT a D. 2 independent and 1 Safeguard WITH a 7.3 second Monticello NRC Written Exam Page: 46 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 16 ANSWER EXPLANATION Answer: A DISTRACTOR ANALYSIS A. Correct: The applicant must have knowledge of the reactor water low level inputs to the RPS logic. Per B.05.06-02, for the RPS, Reactor vessel level is sensed by four level transmitters, LT-2-3-57A, LT-2-3-57B, LT-2-3-58A and LT-2-3-58B. The reactor low water level scram setpoint for RPS is +9 inches. The same level transmitters are used for the Low-Low Water Level Primary Containment Isolation system to isolate Group I and III valves at -47 inches. For the ATWS system, Level is sensed by four different level transmitters (LT-2-3-180A, LT-2-3-180B, LT-2-3-180C, and LT-2-3-180 D). An ATWS system trip is initiated at -47 inches. To prevent the ATWS trip on low-low water level from affecting the ECCS system performance, a 7.3-second time delay is provided for this trip. If the low-low level condition clears before the time delay relay times out, the relay will reset and the ATWS trip will not occur. Per B.05.07-02, for Reactor Level Control, Reactor Water level is measured by two independent level transmitters (LT-6-52A and LT-6-52B) and a third level input is obtained from the B Safeguards level transmitter LT-2-3-85A.

B. Incorrect: The ATWS trip system employs a 7.3 second time delay so ECCS system performance is not affected. Plausible the applicant confuses the ATWS and RPS logic.

C. Incorrect: Per M-8107L-046 (Reactor Level Control) Rev. 12, Reactor Level Control level instrumentation includes 2 independent level transmitters (LT-6-52A and B) and a third level input is received from Safeguards transmitter (LT-2-3-85A). Plausible the applicant confuses the LT from Reactor Level Control and RPS.

D. Incorrect: plausible the applicant confuses RPS level input characteristics with the ATWS and Reactor Level Control systems.

Monticello NRC Written Exam Page: 47 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 16 QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (295031 Reactor Low Water Level / 2) EK2.11 Knowledge of the interrelations between REACTOR LOW WATER LEVEL and the following:

(CFR: 41.7 / 45.8) Reactor protection system Importance Rating: 4.4 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the interrelations between Reactor Low Water Level and the Reactor Protection System; specifically testing the applicants knowledge of the low water level inputs to RPS.

SRO Justification: N/A Technical

References:

Ops Man B.05.06-02 and -03, B.05.07-02, ARP 5-B-19, M-8107L-072 (Reactor Protection System) Rev. 16 and M-8107L-046 (Reactor Level Control) Rev. 12 Proposed references to be provided: None Learning Objective: 6 of M-8107L-072 (Reactor Protection System) Rev. 16 Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History: Used Questions 875 and 1227 from BANK as starting points but this question is significantly different Comments:

Monticello NRC Written Exam Page: 48 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 17 Points: 1.0 An ATWS has occurred with the following conditions:

• Boron was injected using SBLC

• A Blowdown is being performed and C.5-3303 (ALT DEPRESSURIZATION) was entered Which C.5-3303 RWCU Depressurization Modes can be used once Boron is injected?

1. Heat Reject Mode with Filter/Demins bypassed and the SBLC Isolation bypassed

2. Dump Flow Mode to the hotwell or radwaste A. 1 ONLY B. 2 ONLY C. 1 AND 2 D. NEITHER Monticello NRC Written Exam Page: 49 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 17 ANSWER EXPLANATION Answer: A DISTRACTOR ANALYSIS With ATWS conditions and boron injected into the RPV the alternate depressurization systems allows use of the RWCU normal mode as it is OK to defeat isolations including SBLC. The RWCU dump mode is not allowed with boron injected into the core as this would result in the loss of boron from the RPV.

A. Correct, the normal mode is allowed B. Incorrect, the dump mode would result in loss of boron from the RPV C. Incorrect, the dump mode would result in loss of boron from the RPV D. Incorrect, as stated above the normal mode is allowed QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 295037 SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown.

EK1.05 Knowledge of the operational implications of the following concepts as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : Cold shutdown boron weight: Plant-Specific Importance Rating: 3.4 10 CFR Part 55: 55.41.8 to 10 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant identify that alternate depressurization is not permitted with the RWCU dump mode, as this would result in the loss of boron from the RPV.

SRO Justification: N/A Technical

References:

C.5-2002 Ops Man C.5.1-2002 Monticello NRC Written Exam Page: 50 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 17 Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank X Bank _

Question History: MNGP Bank question 2002 Comments:

Monticello NRC Written Exam Page: 51 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 18 Points: 1.0 The plant was operating at rated conditions when a steam line rupture occurred in the Turbine Building (TB). The following timeline of events occurred:

Time Event 1300 The plant is manually scrammed 1303 Main condenser vacuum is degrading 1305 07-B-11 (VACUUM 20 IN TRIP #1) is received 1310 V-AH-1/2 (TB AIR HANDLING UNITS) automatically start 1312 03-B-55 (REACTOR BLDG EXH PLENUM HI RAD) is received 1315 The Standby Gas Treatment (SBGT) System automatically starts.

Complete the following statement that describes the status of V-AH-1/2 at 1320 and why?

V-AH-1 & 2 will A. trip because of rising pressure in the TB.

B. trip because of the rising D/P in the RB Plenum Room.

C. remain running because of the high TB temperature condition.

D. remain running to prevent an unmonitored release from the TB.

Monticello NRC Written Exam Page: 52 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 18 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS The controls of the Turbine Building air handling units V-AH-1 and V-AH-2 are interlocked with a differential pressure switch (dPS-9299), so that when the Plenum Room pressure increases to -.2 inch WC the air handling units will shut off and the Plenum Room ceiling dampers will close. A time delay of one minute is provided before an isolation associated with this pressure increase takes place (62-9299 located in Panel N-9302). Automatic initiation of SBGT on high radiation in the exhaust plenum (26 mrem/hr) and the associated trip of V-EF-20, V-EF-21 and V-EF-22 will cause the Plenum Room pressure to become slightly positive and therefore trip V-AH-1 and V-AH-2.

A is incorrect: Although TB air pressure will change when vacuum degrades, this will not trip V-AH-1 & 2.

C is incorrect: These units auto start on 250°F in the TB but they will trip approximately 1 minute after SBGT starts.

D is incorrect: Plausible to remain running for release purposes but they will trip.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (295038 High Off-site Release Rate / 9) EK2.03 Knowledge of the interrelations between HIGH OFF-SITE RELEASE RATE and the following:

(CFR: 41.7 / 45.8) Plant ventilation systems Importance Rating: 3.6 10 CFR Part 55: 41.9 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the interrelations between high off-site release rate and plant ventilation systems; specifically testing the applicants knowledge of non-Control Room ventilation systems response to an accident condition.

Monticello NRC Written Exam Page: 53 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 18 SRO Justification: N/A Technical

References:

B.08.07-02/03 Proposed references to be provided: None Learning Objective: 5.l of M-8107L-027 (HVAC) Rev. 20 Cognitive Level: Higher X Lower _

Question Source: New Modified Bank _

Bank X Question History: ILT Bank - 2015 Audit Exam Comments: Modified distracter A for plausibility.

Monticello NRC Written Exam Page: 54 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 19 Points: 1.0 The plant was operating at rated conditions in a normal electric plant lineup. An event occurs that results in the following:

• Annunciator 20-A-08 (FIRE) is alarming

• Annunciator 8-A-01 (NO. 1 GENERATOR LOCKOUT) is alarming

• The Main Transformer Fire System status is shown below The BOP operator rotates the arming collar and depresses the pushbutton shown above.

Which of the following, if any, is a valid reason for performing these actions?

A. These actions MANUALLY INITIATE deluge spray on the Main Generator Transformer.

B. These actions MANUALLY TRIP open the 345 KV Main Generator output breakers 8N7 and 8N8.

C. These actions MANUALLY TRIP open the Main Generator Field breaker which results in a Generator LOCKOUT.

D. These actions are NOT required; the Main Transformer Deluge IS currently spraying water on the transformer.

Monticello NRC Written Exam Page: 55 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 19 ANSWER EXPLANATION Answer: A DISTRACTOR ANALYSIS The deluge for the Main XFMR will not automatically spray when the LHD detects a fire. The Detector Operated light indicates that a fire has been detected which auto opens the deluge inlet valve which fills the piping with water. In order to actually spray water, the Main XFMR must be verified dead (Main Generator LOCKOUT) then the arming collar must be turned and the pushbutton pushed. This action is called for in the abnormal procedure for FIRE and ARP 20-A-08.

B and C are incorrect: The Main Generator Lockout will automatically trip these breakers.

Plausible that this action must be manually performed for a fire.

D is incorrect: Plausible to think this action isnt required because the turbine building siding deluge automatically sprays when the detector is operated. The Main XFMR deluge spray doesnt auto initiate as indicated by the System Operated light being off.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 600000 Plant Fire On Site AK3.04 Knowledge of the reasons for the following responses as they apply to PLANT FIRE ON SITE: Actions contained in the abnormal procedure for plant fire on site Importance Rating: 2.8 10 CFR Part 55: 55.41.10 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the knowledge of why certain actions are taken in the control room for a plant fire on site.

SRO Justification: N/A Technical

References:

Ops Man C.4-B.08.05.A ARP 20-A-08 Monticello NRC Written Exam Page: 56 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 19 Proposed references None to be provided:

Learning Objective: M-8107L-010, LO 7, 9 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: 2015 NRC Exam Comments:

Monticello NRC Written Exam Page: 57 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 20 Points: 1.0 The plant is at rated conditions with the Automatic Voltage Regulator (AVR) in AUTOMATIC. A grid disturbance resulted in a variance in grid voltage and frequency and an entry into C.4-F (RAPID POWER REDUCTION). The following conditions now exist:

• Main Generator is ON-LINE with the following parameters:

o 450 MWe o 250 MVAR o H 2 pressure 30 psig o Frequency steady at 59.6 Hz

• Annunciator 8-A-34 (1-BUS 345 KV LOW FREQ) is in alarm (Alarm response procedure attached).

The Transmission System Operator (TSO) requests that MNGP raise their output to help improve grid stability.

Which one of the following represents the highest load increase that is within the limits of the attached Reactive Capability vs. Megawatt Load curves?

A. Load cannot be increased; an immediate reactor shutdown and turbine trip is required.

B. Raise Generator output to 600 MWe Maintain Generator VAR Load at 250 MVAR C. Raise Generator output to 600 MWe Lower Generator VAR load to 150 MVAR D. Raise Generator output to 700 MWE Lower Generator VAR Load to 150 MVAR Monticello NRC Written Exam Page: 58 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 20 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS Grid disturbances can affect the nominal generator frequency of 60 Hz. Annunciator 8-A-34 will alarm at 59.8 Hz. Applicant must understand the nominal characteristics of generator operation and the generator set points. Applicant should recognize that the above initial conditions are within operating limits and do not require any immediate actions. ARP 8-A-34 directs raising generator load as directed by the TSO as long as the generator is operated within limits.

A. Incorrect; per ARP 8-A-34 (1-BUS 345 KV LOW FREQ) the generator may run indefinitely at this frequency and the current load is within the limits for generator gas pressure of 30 psig.

B. Incorrect; MW and MVAR are outside the limits for generator gas pressure of 30 psig; plausible because load is within the limits for 45 psig.

C. Correct; MW and MVAR are within the limits for generator gas pressure of 30 psig.

D. Incorrect; MW and MVAR are outside the limits for generator gas pressure of 30 psig; plausible because load is within the limits for 45 psig.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (700000 Generator Voltage and Electric Grid Disturbances / 6) AA1.01 Ability to operate and/or monitor the following as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: (CFR: 41.5 and 41.10 / 45.5, 45.7, and 45.8)

Grid frequency and voltage Importance Rating: 3.6 10 CFR Part 55: 41.5 and 41.10 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants ability to respond to a grid disturbance and their knowledge of the normal and abnormal operating characteristics of the generator and associated equipment, specifically frequency and voltage Monticello NRC Written Exam Page: 59 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 20 SRO Justification: N/A Technical

References:

Ops Man B.09.02-02, -05, and -06, ARP 8-A-34, 8-A-11, 8-11-31, M-8107L-034 (Generation) Rev. 17 Proposed references to be provided: 8-A-34 (1-BUS 345 KV LOW FREQ) &Figure 2 of Ops Man B.09.02-02 Learning Objective: LO 8.c of M-8107L-034 (Generation) Rev. 17 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments: Question developed from MNGP 2009 NRC Exam Question 20 and MNGP Exam BANK Question 716.

Monticello NRC Written Exam Page: 60 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 21 Points: 1.0 At 1400 the plant was at rated conditions when the following annunciators alarmed.

• OFF GAS HI RADIATION (4-A-12)

• OFF-GAS TIMER ACTUATED (4-A-2)

Complete the statements below?

The earliest time off-gas flow will be automatically terminated is __(1)__

Assuming no operator actions, the reactor will automatically scram __(2)__ the turbine trip on low condenser vacuum.

A. (1) 1415 (2) before B. (1) 1415 (2) after C. (1) 1430 (2) before D. (1) 1430 (2) after Monticello NRC Written Exam Page: 61 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 21 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS The off-gas isolation occurs after 30 minutes. The reactor scram setpoint is 22.3 Hg Vac, the turbine trip setpoint is 20 Hg Vac A. (1) Incorrect (2) Correct B. (1) Incorrect (2) Incorrect C. (1) Correct (2) Correct D. (1) Correct (2) Incorrect QUESTION BACKGROUND Tier: 1 Group: 2 K/A: 295002 Loss of Main Condenser Vac AA2.02 Ability to determine and/or interpret the following as they apply to LOSS OF MAIN CONDENSER VACUUM : Reactor power: Plant-Specific Importance Rating: 3.2 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine that the alarm condition will result in automatic isolation of the offgas system following a 30 minute time delay, and once offgas is isolated the condenser vacuum will degrade to the reactor auto scram setpoint which occurs at higher vacuum then the turbine trip.

Monticello NRC Written Exam Page: 62 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 21 SRO Justification: N/A Technical

References:

Ops Man C.4-B.06.03.A C.6-004-A-02 Proposed references None to be provided:

Learning Objective:

Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank X Bank _

Question History: MNGP exam bank 1366 Comments:

Monticello NRC Written Exam Page: 63 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 22 Points: 1.0 The plant is operating in MODE 1 when a small break LOCA concurrent with a Loss of Offsite Power occurs. Given the following current conditions:

• Control rods are fully inserted

• Both EDGs are operating and supplying their respective busses

• The only system injecting to the vessel is CRD

• Reactor pressure is 470 psig and lowering at 10 psig/min

• DW pressure is 0.6 psig and rising at 0.05 psig/min

• RPV level is -47 inches and lowering at 0.5 inches/min What is the status of the Core Spray system 17 minutes later?

A. Core Spray pumps are not running.

B. Core Spray pumps are running and injecting to the RPV.

C. Core Spray pumps are running but the injection valves have not opened.

D. Core Spray pumps are running injection valves are open, but no injection is occurring.

Monticello NRC Written Exam Page: 64 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 22 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A. Incorrect; CS will start and injection valves open when RPV pressure is <420 psig (at T+5 min). [Outboard injection valves are normally open, inboard injection valves are normally closed, and open when Reactor pressure is below valve pressure interlock setpoint.]

B. Correct; CS will start and injection valves open when RPV pressure is <420 psig (at T+5 min). The ADS timer will actuate when approximately two minutes later (T=15 minutes) but the SRVs wont open until T=16:47. Injection will occur a short time later when RPV press falls below the discharge head of the pump, less static head and dynamic head losses -

approximately 310 psig. Specifically, an automatic initiation signal along with <420 psig Rx pressure for CS injection to occur.

Core Spray is initiated when ONE of the following conditions has been met:

a. High drywell pressure (>1.84 psig)

OR

b. Reactor low-low water level AND

1) Reactor pressure < 420 psig OR

2) RPV level < -47 for 15 minutes In addition, the Inboard and Outboard Isolation/Injection valves (MO-1753 or MO-1754 and MO-1751 or MO-1752) also receive an open demand at <420 psig (assuming one of the above conditions are met).

C. Incorrect; See B above.

D. Incorrect; See B; plausible if examinee does not calculate pressure drop correctly QUESTION BACKGROUND Tier: 1 Group: 2 K/A: (295007 High Reactor Pressure / 3) AK2.04 Knowledge of the interrelations between HIGH REACTOR PRESSURE and the following: (CFR: 41.7 / 45.8)

LPCS Importance Rating: 3.2 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A Monticello NRC Written Exam Page: 65 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 22 K/A Match: K/A is matched by testing the applicants knowledge of the Low Pressure Core Spray Systems inability to inject at high reactor pressure; the LPCS pumps will only inject once reactor pressure drops below a certain threshold and auto initiation signals are met.

SRO Justification: N/A Technical

References:

Ops Man B.03.01-01 and -02, B.03.03-01 and -02, and M-8107L-005 (Core Spray) Rev. 18 Proposed references to be provided: None Learning Objective: 5.d of M-8107L-005 (Core Spray) Rev. 18 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Modified from MNGP Exam BANK Question 138 (M8107L-005 7)

Comments:

Monticello NRC Written Exam Page: 66 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 23 Points: 1.0 The plant was at rated conditions when a transient resulted in the following conditions:

• Bus 15 is unavailable

• Bus 16 is being supplied by 12 EDG

• Torus water temperature is 112°F and rising

• Torus Cooling is in service using 12 & 14 RHR pumps and 14 RHRSW pump

• A RHR system flow is 0 gpm

• B RHR system flow is 8000 gpm

• B RHRSW system flow is 3500 gpm By procedure, which of the following actions would be allowed to be performed to increase the heat removed from the torus?

A. Start #12 RHRSW pump.

B. Throttle OPEN MO-2003 (HX BYPASS).

C. Throttle OPEN CV-1729 (RHR HX SW OUTLET).

D. Throttle OPEN MO-2009 (TORUS COOLING INJ/TEST INBOARD).

Monticello NRC Written Exam Page: 67 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 23 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS IAW Operations Manual B.03.04-05: THROTTLE OPEN MO-2009, to provide 4000 gpm per pump OR Adjust flow for approximately 3500 gpm per pump using CV-1729 OR CLOSE MO-2003 HX Bypass.

CAUTION Due to Emergency Diesel Generator transient loading capability, a single EDG may only provide power to three large motors (RHRSW, Core Spray, or RHR) at one time.

In accordance with Operations Manual B.08.01.03-05:

If emergency or abnormal conditions require, a single RHRSW pump may operate at flow rates up to 6000 gpm. Operation at flows greater than this may result in damage to the pump thrust bearing A. Incorrect, three large pumps are already running on the EDG.

B. Incorrect, opening the heat exchanger bypass valve reduce cooling of the torus water C. Correct, a single RHRSW pump may operate at up to 6000 gpm during abnormal or emergency conditions D. Incorrect, the RHR is at the pump limit of 4000 gpm per pump QUESTION BACKGROUND Tier: 1 Group: 2 K/A: 295013 High Suppression Pool Temp.

AA1.01 Ability to operate and/or monitor the following as they apply to HIGH SUPPRESSION POOL TEMPERATURE : Suppression pool cooling...

Importance Rating: 3.9 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine a method of increasing cooling to the torus water that is permitted by procedures given a set of plant conditions.

Monticello NRC Written Exam Page: 68 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 23 SRO Justification: N/A Technical

References:

Operations Manual B.03.04-05 Operations Manual B.08.01.03-05 Proposed references None to be provided:

Learning Objective: M-8107L-023 LO 9 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: MNGP exam bank 466 Comments:

Monticello NRC Written Exam Page: 69 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 24 Points: 1.0 The reactor was operating at rated conditions when an event occurred. Torus level is reported by the BOP operator to be +3.9 feet.

Complete the following statement concerning the current torus water level?

At this torus water level A. HPCI operation is not permitted.

B. SBGT cannot be used to vent non-condensable from the torus airspace.

C. SRV operation could cause SRV tail pipe damage resulting in containment failure.

D. Vacuum Breakers will not function to relieve non-condensable from the torus to the drywell.

Monticello NRC Written Exam Page: 70 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 24 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A, B, and D are incorrect but plausible variations since each system deals with pressure relief/suppression pool level control. Per C.5.1-1200, A. Incorrect: HPCI is a concern for lower suppression pool level (not high); if suppression pool level falls to -3.7 feet, then the HPCI exhaust will uncover but operation can still be permitted under extreme conditions.

B. Incorrect: SBGT can be used to vent the torus at levels up to 11.3 ft.

C. Correct: A torus water level of 3.7 feet corresponds to the Maximum Pressure Suppression Primary Containment Water Level, which bounds both the Pressure Suppression Pressure and the SRV Tail Pipe Level Limit. SRV operation with torus water level above the SRV Tail Pipe Level Limit could damage the SRV discharge lines. This, in turn, could lead to containment failure from direct pressurization and damage to equipment inside the containment from pipe-whip and jet-impingement loads. At Monticello, the SRV Tail Pipe Level Limit is bounded by the Maximum Pressure Suppression Primary Containment Water Level for all RPV pressures below approximately 1150 psig D. Incorrect: Vacuum Breakers will operate under high suppression level up to +4.2 feet; above

+4.2 feet the Vacuum Breakers may not function as designed.

QUESTION BACKGROUND Tier: 1 Group: 2 K/A: (295029 High Suppression Pool Wtr Lvl / 5) EK1.01 Knowledge of the operational implications of the following concepts as they apply to HIGH SUPPRESSION POOL WATER LEVEL: (CFR: 41.8 to 41.10) Containment integrity Importance Rating: 3.4 10 CFR Part 55: 41.8 to 41.10 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the high suppression pool water levels operational impact on the primary containments structural integrity.

Monticello NRC Written Exam Page: 71 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 24 SRO Justification: N/A Technical

References:

EOP C.5-1200, C.5.1-1200 Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank _

Bank X Question History: Question 26 from 2010 MNGP ILT NRC Comments:

Monticello NRC Written Exam Page: 72 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 25 Points: 1.0 The plant was at rated conditions when a transient occurred. Given the following:

• RWCU area temperature is 175°F

• HPCI area temperature is 180°F Complete the following statements.

An automatic isolation of the __(1)__ system has occurred.

The associated PCIVs close to __(2)__.

A. (1) HPCI (2) isolate the HPCI steam line ONLY B. (1) RWCU (2) isolate the RWCU supply and return lines ONLY C. (1) HPCI (2) isolate the HPCI steam line AND Torus suction line D. (1) RWCU (2) isolate the RWCU supply and return lines AND the Recirc Sample line Monticello NRC Written Exam Page: 73 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 25 ANSWER EXPLANATIONs Answer: D DISTRACTOR ANALYSIS RWCU Isolation valves and the Recirc sample line isolation valves close when 170°F is reached in the RWCU room. The examinee must recognize the setpoint and which valves close for a Group 3 isolation.

A. (1) Incorrect, the HPCI system does not isolate until 187.5°F (2) Incorrect, when HPCI isolates both the steam line and torus suction line isolate.

B. (1) Correct, An isolation of the RWCU system is generated when room temperature exceeds 170°F (2) Incorrect, both the RWCU lines and Recirc sample line are isolated.

C. (1) Incorrect, the HPCI system does not isolate until 187.5°F (2) Correct, when HPCI isolates both the steam line and torus suction line isolate.

D. (1) Correct, An isolation of the RWCU system is generated when room temperature exceeds 170°F (2) Correct, both the RWCU lines and Recirc sample line are isolated.

QUESTION BACKGROUND Tier: 1 Group: 2 K/A: 295032 High Secondary Containment Area Temperature EK3.03 Knowledge of the reasons for the following responses as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE : Isolating affected systems...

Importance Rating: 3.8 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine based on plant conditions the correct PCIS isolation and why multiple systems are isolated. Specifically, the Recirc Sample line isolated because this piping runs through the RWCU room.

Monticello NRC Written Exam Page: 74 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 25 SRO Justification: N/A Technical

References:

Ops Man B.05.06-02 Ops Man B.03.02-01 Proposed references None to be provided:

Learning Objective: M-8107L-023 LO 9 Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 75 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 26 Points: 1.0 Which of the following describes the response to an UPSCALE trip of a single Reactor Building Exhaust Plenum radiation monitor?

The UPSCALE trip will.

1) Isolate the normal Reactor Building ventilation exhaust path

2) Initiate operation of the Standby Gas Treatment System

3) Close the Primary Containment Purge and Vent valves

4) NOT result in any of the above conditions A. 1 ONLY B. 1 and 2 ONLY C. 1 and 2 and 3 D. 4 Monticello NRC Written Exam Page: 76 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 26 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A, B, and D are incorrect but plausible variations of the actions of the Reactor Building Ventilation system upon receiving a high radiation (upscale signal). The applicant must first recognize that a single upscale in the Reactor Building exhaust plenum (or the fuel pool monitors) trip will satisfy the trip circuit logic (one upscale or TWO downscale satisfy the circuit) - thus D is plausible since the applicant may believe more than 1 upscale is required. In addition, the applicant must know the actions when the trip circuit is satisfied.

C is correct: In accordance with Ops Manual B.05.11-02, there are two Reactor Building exhaust vent plenum monitors (RM-17-452A and RM-17-452B), and one upscale or two downscale will initiate the following:

a. Reactor Building Ventilation System shutdown

b. Standby Gas Treatment System initiation

c. Primary Containment purge and ventilation valves closure (select Group II Primary Containment valves)

Annunciator B-3-55 (Reactor Building Exhaust High Rad) will alarm when output of either channel exceeds 20 mrem/hr.

Monticello NRC Written Exam Page: 77 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Tier: 1 Group: 2 K/A: (295034 Secondary Containment Ventilation High Radiation / 9) 2.1.28 Knowledge of the purpose and function of major system components and controls. (CFR: 41.7)

Importance Rating: 4.1 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the function of the major components within Secondary Containment Ventilation associated with a high radiation signal.

SRO Justification: N/A Technical

References:

Ops Manual B.05.11-01 and -02, and M-8107L-077 (Process Radiation Monitoring and Area Radiation Monitors) Rev. 15 Proposed references to be provided: None Learning Objective: 2.j of M-8107L-077 (Process Radiation Monitoring and Area Radiation Monitors) Rev. 15 Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History: N/A Comments:

Monticello NRC Written Exam Page: 78 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 27 Points: 1.0 The plant was at rated conditions with HPCI in a normal standby lineup. Given the following:

• A HPCI suction line break occurs

• The HPCI area water level is rising

• The CRS has entered C.5-1300 (SECONDARY CONTAINMENT CONTROL).

Complete the statements below?

The source of the leak is from the __(1)__ .

When the HPCI area reaches MAX SAFE, __(2)__ .

A. (1) CSTs (2) a reactor scram must be inserted B. (1) Torus (2) a reactor scram must be inserted C. (1) CSTs (2) the plant can remain at power since the leak is not from a PRIMARY system D. (1) Torus (2) the plant can remain at power since the leak is not from a PRIMARY system Monticello NRC Written Exam Page: 79 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 27 ANSWER EXPLANATIONs Answer: C DISTRACTOR ANALYSIS The normal suction supply for HPCI is from the CSTs. The CST is not a source of primary water discharging into the reactor building so the operator will not be directed to shut down until 2 or more areas are above max safe values for the same parameter.

A. (1) Correct (2) Incorrect B. (1) Incorrect (2) Incorrect C. (1) Correct (2) Correct D. (1) Incorrect (2) Correct QUESTION BACKGROUND Tier: 1 Group: 2 K/A: 295036 Secondary Containment High Sump/Area Water Level 2.4.6 Knowledge of EOP mitigation strategies.

Importance Rating: 3.7 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine if the plant conditions require EOP mitigate action in the water levels leg. The required action to mitigate the transient is dependent on the source of discharge.

Monticello NRC Written Exam Page: 80 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 27 SRO Justification: N/A Technical

References:

Ops Man C.5-1300, Secondary Containment Control Ops Man C.5.1-1300, Secondary Containment Control Proposed references to be provided: None Learning Objective: M-8107L-023 LO 9 Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 81 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 28 Points: 1.0 A small break LOCA has occurred in the Drywell. RPV pressure control is on LO LO-SET and the following annunciators are in alarm:

Panel C-03B Panel C-05B Based ONLY on these alarms, what is the expected status of the RHR system?

A. A LPCI initiation signal is present AND RHR pumps are injecting into the RPV B. A LPCI initiation signal is present but water is NOT injecting into the RPV because the LPCI OUTBOARD injection valves are still closed C. A LPCI initiation signal is present but water is NOT injecting into the RPV because the LPCI INBOARD injection valves are still closed.

D. A LPCI initiation signal is NOT present therefore RHR pumps are NOT injecting water into the RPV.

Monticello NRC Written Exam Page: 82 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 28 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A, B, and D are incorrect but plausible variations of a LPCI initiation signal and injection: The applicant must have knowledge of the set-points associated with the listed annunciators and how they correspond to LPCI injection. A LPCI initiation signal AND RPV pressure below 420 psig must be present before the Inboard and/or Outboard Injection valves will open automatically.

A LPCI Initiation Signal is present as indicated by Annunciator 3-B-53 (High DW Pressure).

Annunciator 5-B-24 may be associated with RPV water level being below normal (less than +30),

but since annunciator 3-A-38 (Low-Low Level) is not stated to be active, the LPCI Initiation signal on low-low RPV water level is not present. Annunciator 3-A-30 (Reactor Low Pressure (420 psig))

would not active since pressure control is on LL-SET, therefore the permissive to open the injection valves is not met and neither LPCI loop will be injecting.

A. Incorrect; A LPCI Injection signal is present, but RPV pressure is above 420 psig as is evident by the absence of annunciator 3-A-30. The Inboard Injection valves will not be open.

B. Incorrect; A LPCI Injection signal is present, but RPV pressure is above 420 psig as is evident by the absence of annunciator 3-A-30. The Inboard Injection valves will not be open. The Outboard Injection valves are NORMALLY open.

C. Correct: A LPCI Injection signal is present, but RPV pressure is above 420 psig as is evident by the absence of annunciator 3-A-30. The Inboard Injection valves will not be open. The Outboard Injection valves are NORMALLY open.

D. Incorrect; A LPCI Injection signal is present as evidenced by active Annunciator 3-B-53 (High DW Pressure)

Monticello NRC Written Exam Page: 83 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 28 QUESTION BACKGROUND Tier: 1 Group: 2 K/A: (203000 RHR/LPCI: Injection Mode) 2.4.31 Knowledge of annunciator alarms, indications, or response procedures. (CFR: 41.10 / 45.3)

Importance Rating: 4.2 10 CFR Part 55: 41.10 / 45.3 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the annunciator alarms, and their corresponding setpoints, associated with the RHR/LPCI injection mode SRO Justification: N/A Technical

References:

Ops Man B.03.04-01 and -03, and M-8107L-023 (RHR System) Rev. 20 Proposed references to be provided: None Learning Objective: 7 of M-8107L-023 (RHR System) Rev. 20 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Question 502 from BANK and Question 56 of 2003 MNGP ILT - modified question stem with the replacement of metric with annunciators in order to match K/A and changed a distractor Comments: This question has ties to Question 502 in the exam bank, but the question stem has been modified significantly and 1 distractor has been changed so it likely qualifies as a significantly modified BANK question Monticello NRC Written Exam Page: 84 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 29 Points: 1.0 The plant was operating in Mode 2 with a startup in progress. A reactor coolant leak developed in the Reactor Vessel Bottom Head and following parameters were recorded at the indicated times:

TIME Reactor Level Reactor Pressure Drywell pressure 00:00:00 +30 inches 900 psig 0.5 psig 00:05:00 +8 inches 850 psig 0.7 psig 00:10:00 -15 inches 750 psig 1.1 psig 00:20:00 -48 inches 600 psig 1.9 psig 00:30:00 +30 inches 400 psig 2.4 psig Based on the above timeline, which of the following is correct?

A. ONLY a Partial Group 2 Isolation signal has occurred by 00:05:00.

B. ALL Main Steam Isolation Valves are closed at 00:10:00.

C. Division 1 RHR was selected for LPCI Loop injection path by 00:20:00.

D. Division 2 RHR LPCI injection valves have received an open signal by 00:30:00.

Monticello NRC Written Exam Page: 85 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 29 ANSWER EXPLANATION ANSWER D DISTRACTOR ANALYSIS A. Incorrect: A Full Group 2 Isolation would occur at this time.

B. Incorrect: With the plant in Mode 2 the Mode Switch would be in STARTUP. Therefore the <840 psig in RUN Group 1 isolation will be bypassed.

C. Incorrect: Plausible as LPCI loop selection takes place at -47 and 1.84 psig, Division 2 is the default loop for LPCI injection and since the leak is in the bottom head, Division 1 would NOT be selected.

D. Correct: LPCI loop selection takes place at -47 and 1.84 psig, division 2 is the default loop for LPCI injection and since RCS pressure is <420 psig and drywell pressure is

>1.84 psig the division 2 RHR injection valves will be demanded open.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 203000 RHR/LPCI: Injection Mode A4.08 Ability to manually operate and/or monitor in the control room: Reactor pressure Importance Rating: 4.3 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: The applicant is given parameters that would be monitored in the control room that affect the RHR system lineup to LPCI. Based on the plant parameters LPCI injection via the Division 2 RHR system would occur since RCS pressure has dropped low enough to allow the injection valves to open.

SRO Justification: N/A Monticello NRC Written Exam Page: 86 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 29 Technical

References:

B.03.04-01, B.03.04-02 Proposed references to be provided: None Learning Objective: M-81 07L-023, RHR System, 7. Describe the RHR system operation, controls, automatic functions, alarms, setpoints, and or interlocks.

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Significantly modified from MNGP NRC Exam 2015, Q 11 Comments:

Monticello NRC Written Exam Page: 87 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 30 Points: 1.0 The plant is in MODE 4 with the 12 RHR Pump operating in a Normal Shutdown Cooling Mode.

The desired RCS temperature control band is 120-140°F.

• Both Reactor Recirculation pumps are OFF

• The CRD and RWCU systems are SHUTDOWN Time Recirc Loop A Recirc Loop B 12 RHR HX Inlet RWCU Inlet Suction Suction Temperature Temperature Temperature Temperature 0900 141 139 140 139 0915 140 139 138 140 0930 140 138 137 139 0945 139 139 135 139 1000 138 138 134 140 Give the above information, what is the RCS Heatup/Cooldown rate.

A. Cooling down at 3°F/hr B. Cooling down at 1°F/hr C. Cooling down at 6°F/hr D. Heating up at 1°F/hr Monticello NRC Written Exam Page: 88 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 30 Answer Explanation Answer: C DISTRACTOR ANALYSIS A, B, and D are incorrect but plausible variations of actions in order to decrease the recirculation loop temperature at a slower rate (i.e., decrease cooldown rate).

A. Incorrect: There may be some back-flow through the 11 RR Loop, but the temperature of the loop will not be representative of the RCS temperature.

B. Incorrect: With the 12 RR Loop suction or discharge valve shut there will be no flow through the loop and the indicated temperature will not be representative of the RCS temperature.

C. Correct: The coolant flowing throught the in-service RHR train will be representative of the RCS temperature.

D. Incorrect: With the RWCU system shutdown there will be no flow through the system and the indicated temperatue will not be representative of the RCS temperature.

Question Background Tier: 1 Group: 2 K/A: (205000 Shutdown Cooling) K1.03 Knowledge of the physical connections and/or cause-effect relationships between SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and the following: Recirculation loop temperature Importance Rating: 3.4 10 CFR Part 55: CFR: 41.2 to 41.9 / 45.7 to 45.8 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the cause-effect relationships between the SDC system and the recirculation loop temperature, specifically focusing on how altering SDC components effect recirculation loop temperature SRO Justification: N/A Monticello NRC Written Exam Page: 89 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 30 Technical

References:

Ops Man B.03.04-01 and -02, Startup Procedure B.03.04-05.D.2, and M-8107L-023 (RHR System) Rev. 20 Proposed references to be provided: None Learning Objective: 7 of M-8107L-023 (RHR System) Rev. 20 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History: Question 29 of 2010 MNGP ILT - significantly modified to match K/A Comments: This question has ties to Question 29 from the 2010 MNGP ILT, but it has been modified significantly; thus it may qualify as a NEW question.

Also, the question could be potentially be strengthened by including more information in the question stem (i.e., including a list of times and temperatures, like what is done in Question 514 in the Exam BANK - see the comments in the top of this document).

Monticello NRC Written Exam Page: 90 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 31 Points: 1.0 The reactor was at rated conditions when a Loss of Division 2, 125 VDC occurred.

Which of the following statements describes the effect this will have on the plant?

A. The 12 EDG will NOT start automatically.

B. The 12 EDG WILL start automatically and the field WILL flash.

C. The HPCI Automatic Initiation Logic will become NON-Functional.

D. MO-2034 (HPCI STEAM LINE INBOARD ISO) can NOT be closed from C-03.

Monticello NRC Written Exam Page: 91 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 31 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS The EDG will auto start from the fast start relay for the 11 EDG which is powered from division 1 125 VDC. The field flash circuitry, however, is not cross connected and would require jumpers to accomplish this function. Division 2, 125 VDC powers D21, Circuit 15 (via C39) supplies HPCI logic power (for initiation, trip, and Group IV isolation).

A. Incorrect, see above.

B. Incorrect, see above.

C. Correct, see above.

D. Incorrect, Control power for the only AC-powered HPCI valve (MO-2034) is 120 Vac supplied directly from the control power transformer in that MCC breaker (MCC-143).

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: 206000 HPCI K2.03 Knowledge of electrical power supplies to the following: Initiation logic: BWR-2,3,4 Importance Rating: 2.8 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: The power supply to the initiation logic is lost, the applicant is asked what effect this will have on the plant.

SRO Justification: N/A Technical

References:

Ops Man B.03.02-02 NX-8292012-5 Monticello NRC Written Exam Page: 92 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 31 Proposed references to be provided: None Learning Objective: M-8107L-002 LO 8e Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank X Bank _

Question History: Significantly modified MNGP bank question 796 (M8107L-042 4)

Comments:

Monticello NRC Written Exam Page: 93 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 32 Points: 1.0 The plant is in MODE 1 when a Small Break LOCA occurs. The following is observed:

• Bus 15 is locked-out

• HPCI tripped and is not available

• ADS has automatically initiated

• RPV pressure is 200 psig and lowering

• CS Pump B is operating and injecting into the RPV

• Division 2 RHR pumps are running and aligned for torus cooling

• RPV water level is -55 inches and rising slowly Given the stated conditions, complete the following statement that describes the effect on the ADS System if off-site power is lost to Bus 16:

Automatic Depressurization of the Reactor...

A. continues even if EDG 12 fails to start.

B. will cease, then re-initiate immediately upon restart of the Core Spray pump following restoration of power to Bus 16 from EDG 12.

C. will cease, then re-initiate 107 seconds after a Core Spray pump is restarted following restoration of power to Bus 16 from EDG 12.

D. will cease, then after power is restored to Bus 16 from EDG 12, BOTH inhibit switches A and B must be placed in INHIBIT and back to AUTO to re-establish depressurization.

Monticello NRC Written Exam Page: 94 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 32 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A. Incorrect: ADS will cease since AC interlock logic (K10A and K12A) will de-energize breaking the seal-in (no pumps running).

B. Incorrect: ADS will cease since AC interlock logic (K10A and K12A) will de-energized breaking the seal-in (no pumps running), but the timer will reset and delay initiation for 107 seconds following restart of the CS pump.

C. Correct: When the RHR and CS pumps trip/lose power, relays K10A and K12A are de-energized, which open up the K10A and K12A contacts to de-energize the 107 second timer, closing the valves. If the pumps are restarted, contacts K10A and K12A close to start the107 second timer again.

D. Incorrect: plausible interpretations of how INHIBIT logic works; resetting using inhibit switch is unnecessary.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (209001 LPCS) K3.02 Knowledge of the effect that a loss or malfunction of the LOW PRESSURE CORE SPRAY SYSTEM will have on following: (CFR: 41.7 / 45.4)

ADS logic Importance Rating: 3.8 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of what effect the loss of the CS pumps will have on ADS logic SRO Justification: N/A Monticello NRC Written Exam Page: 95 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 32 Technical

References:

Ops Man B.03.03-01, B.03.03-06 Figure 5, M-8107L-025 (Reactor Pressure Relief) Rev. 23, M-8107L-005 (Core Spray) Rev. 18 Proposed references to be provided: None Learning Objective: 7 from M-8107L-025 (Reactor Pressure Relief) Rev. 23 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Question 43 of 2015 MNGP ILT NRC Written Exam Comments: Potential cueing concern in distractors, as proposed correct answer is the only option that states ADS will continue.

Monticello NRC Written Exam Page: 96 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 33 Points: 1.0 An ATWS event has occurred and the OATC has initiated SBLC from C-05.

Complete the statements below.

Forced circulation is __(1)__ to ensure adequate dispersion of the boron solution into the core.

RWCU pumps __(2)__ receive an automatic trip signal.

A. (1) required (2) will B. (1) required (2) will NOT C. (1) NOT required (2) will D. (1) NOT required (2) will NOT Monticello NRC Written Exam Page: 97 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 33 ANSWER EXPLANATION Answer C DISTRACTOR ANALYSIS A. (1) Incorrect, without a recirculation pump running, natural circulation provides adequate dispersion of the solution into the core (2) Correct B. (1) Incorrect, without a recirculation pump running, natural circulation provides adequate dispersion of the solution into the core (2) Incorrect, RWCU pumps trip on SBLC system actuation unless the isolation interlocks are bypassed per the EOP C. Correct D. (1) Correct (2) Incorrect, RWCU pumps trip on SBLC system actuation unless the isolation interlocks are bypassed per the EOP QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 211000 SLC K4.05 Knowledge of STANDBY LIQUID CONTROL SYSTEM design feature(s) and/or interlocks which provide for the following: Dispersal of boron upon injection into the vessel Importance Rating: 3.4 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: The SBLC system is designed to shutdown the reactor from all conditions at any time in core life, normal mixing via recirculation pumps is not required to ensure borated solution will disperse into the core region, natural circulation is sufficient. The SBLC system provides enough boron solution to meet the shutdown requirements with normal water level including water volume in the RHR shutdown cooling piping and the recirculation loops.

Isolation of RWCU and tripping of the pumps is a start interlock with the actuation of SBLC to ensure borated solution is contained within the RCS volume as described.

Monticello NRC Written Exam Page: 98 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 33 SRO Justification: N/A Technical

References:

Ops Man B.03.05-05 Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History: New Comments:

Monticello NRC Written Exam Page: 99 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 34 Points: 1.0 The reactor was operating at rated conditions when the following occurred:

• BOTH manual scram push buttons were depressed then released.

• Reactor pressure is 750 psig and slowly decreasing.

• The OATC has placed the SCRAM DISCHARGE VOLUME HIGH WATER LEVEL BYPASS switches in BYPASS.

Given the above conditions, which of the following mode switch positions will allow the scram to be reset?

A. SHUTDOWN ONLY B. SHUTDOWN OR REFUEL ONLY C. SHUTDOWN OR START & HOT STBY ONLY D. SHUTDOWN OR REFUEL OR START & HOT STBY ONLY Monticello NRC Written Exam Page: 100 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 34 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A, C, and D are incorrect but plausible interpretations of this RPS logic arrangement: Per the RPS Logic Diagram, although Low Condenser Vacuum / MSIV Closure scram signal can be bypassed in Shutdown OR Refuel OR Start & Hot Stby, Start & Hot Stby is NOT a condition for BYPASS of SDV High Level trip, thus the MS can only be switched to Shutdown OR Refuel to clear ALL the parameters.

B is correct: Reset is possible ONLY if ALL the conditions that caused the trip or scram have been cleared. Per the RPS logic diagram, this can only be accomplished for this RPS logic arrangement if the mode switch is in either SHUTDOWN or REFUEL, Reactor Pressure is <600 psig, AND the SDV high level scram signal key switch is placed in BYPASS (key switch only works with BOTH RPS buses powered) (the latter two conditions are said to apply). In these conditions relays are energized which bypasses the low condenser vacuum, MSIV closure, and SDV Level inputs to RPS.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (212000 RPS) K5.02 Knowledge of the operational implications of the following concepts as they apply to REACTOR PROTECTION SYSTEM: (CFR: 41.5 / 45.3)

Specific Logic arrangements Importance Rating: 3.3 10 CFR Part 55: 41.5 10 CFR 55.43.b  : N/A K/A Match: K/A is matched by testing the applicants knowledge of the logic arrangements of the Reactor Protection System Monticello NRC Written Exam Page: 101 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 34 SRO Justification: N/A Technical

References:

Ops Man B.05.06-06, Figure 6, NX-7834-67-09, (-10), (-11), M-8107L-072 (Reactor Protection System) Rev. 16 Proposed references to be provided: NONE Learning Objective: 6 from M-8107L-072 (Reactor Protection System) Rev. 16 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Question 35 of 2009 MNGP Audit Written Exam Comments:

Monticello NRC Written Exam Page: 102 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 35 Points: 1.0 The plant is in Mode 2 with the following conditions:

• IRM 15 is on Range 3 reading 110/125

• All other IRMs are on Range 3 reading 30-70/125 Subsequently, the high voltage power supply for IRM 15 fails low.

Which IRM trip unit(s) would be actuated by this condition?

1) IRM HI HI

2) IRM INOP

3) IRM DOWNSCALE A. 1 only B. 3 only C. 1 & 2 only D. 2 & 3 only Monticello NRC Written Exam Page: 103 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 35 ANSWER EXPLANATION Answer D DISTRACTOR ANALYSIS A. Incorrect, for IRM HI HI the output of the detector would have to be 119.375 units or more on 0 to 125 unit, a failure of the high voltage supply would cause the output to lower, this is the opposite effect of the malfunction presented B. Incorrect, the IRM downscale is expected for this condition but so is the IRM INOP as presented in answer D.

C. Incorrect, IRM INOP is expected but the IRM HI HI is incorrect as discussed in A.

D. Correct, IRM INOP and IRM DOWNSCALE are expected for High voltage power supply failure QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 215003 IRM K6.04 Knowledge of the effect that a loss or malfunction of the following will have on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM : Detectors Importance Rating: 3.0 10 CFR Part 55: 41.7 10 CFR 55.43.b  : N/A K/A Match: The applicant is asked to determine the effect on IRM system (trip units) when the detector fails due to a loss of high voltage to the detector SRO Justification: N/A Technical

References:

ARP 5-A-5 ARP 5-A-29 Monticello NRC Written Exam Page: 104 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 35 Proposed references to be provided: None Learning Objective: M8107L-064, LO 6 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 105 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 36 Points: 1.0 A reactor startup is in progress when a Control Rod withdraw block is received.

• NO SRMs or IRMs are BYPASSED

• All IRMs are DOWNSCALE on Range 1

• SRM CH 21 is fully INSERTED and reads 8.9 x 104 cps with a 29 second period

• SRM CH 22 is partially WITHDRAWN and reads 2.9 x 102 cps with a 10 second period

• SRM CH 23 is partially WITHDRAWN and reads 9.4 x 101 cps with a 34 second period

• SRM CH 24 is fully INSERTED and reads 6.8 x 100 cps with a 55 second period BYPASSING which of the following will remove the Control Rod withdraw block?

A. SRM CH 21 B. SRM CH 22 C. SRM CH 23 D. SRM CH 24 Monticello NRC Written Exam Page: 106 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 36 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A. Incorrect: plausible to think the applicant doesnt know the UPSCALE setpoint.

B. Incorrect: plausible to think the applicant doesnt know the Retract Not Permitted setpoint.

Also, plausible the applicant thinks a particular period setpoint causes a control rod withdraw block C. Correct: The following SRM trips initiate Control Rod Withdrawal Blocks in the Refuel or Startup Modes/ These rod blocks hold true for each SRM unless it is bypassed. Only one (1)

SRM can be bypassed. These are:

• UPSCALE (>9.07x104 cps (Nominal 1x105 cps)) (bypassed on IRM Range 8 or above) (5-A-12),

• DOWNSCALE (5 cps) (bypassed on IRM Range 3 or above) (5-A-4),

• RETRACT NOT PERMITTED (Detector Not Full In and Channel is <100 cps) (bypassed on IRM Range 3 or above) (5-A-28)

Thus, since SRM CH 23 reads < 100 cps and is NOT FULL IN, a withdrawal rod block is in place for this SRM D. Incorrect: plausible to think the applicant doesnt know the DOWNSCALE setpoint QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (215004 Source Range Monitor) A1.05 Ability to predict and/or monitor changes in parameters associated with operating the SOURCE RANGE MONITOR (SRM)

SYSTEM controls including: (CFR: 41.5 / 45.5) SCRAM, rod block, and period alarm trip setpoints Importance Rating: 3.6 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A Monticello NRC Written Exam Page: 107 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 36 K/A Match: K/A is matched by testing the applicants ability to predict what changes to the SRM system controls would relieve a rod block SRO Justification: N/A Technical

References:

Ops Man B.05.01.01-03, M-8107L-054 (Source Range Monitors) Rev. 10 Proposed references to be provided: None Learning Objective: 6 from M-8107L-054 (Source Range Monitors) Rev. 10 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History: N/A Comments:

Monticello NRC Written Exam Page: 108 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 37 Points: 1.0 A plant startup is in progress with all IRMs on Range 6.

FT-2-110A (REC 11 FLOW XMTR - APRM 1) failed upscale causing the following alarms:

• 5-A-3 (ROD WITHDRAW BLOCK)

• 5-A-30 (APRM FLOW BIAS OFF NORMAL)

Complete the statements below?

__(1)__ is preventing control rods from being withdrawn.

Bypassing APRM 1 __(2)__ allow control rods to be withdrawn A. (1) RBM Flow Comparison >10%

(2) WILL B. (1) RBM Flow Comparison >10%

(2) WILL NOT C. (1) APRM Flow Upscale >110%

(2) WILL D. (1) APRM Flow Upscale >110%

(2) WILL NOT Monticello NRC Written Exam Page: 109 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 37 ANSWER EXPLANATION Answer C DISTRACTOR ANALYSIS The APRM Flow UPSCALE will cause a rod block for the given conditions. The >10% mismatch is an alarm only. A single FT upscale failure only inputs to APRM 1. If APRM 1 is bypassed the rod block will clear.

A. (1) Incorrect (2) Correct B. (1) Incorrect (2) Incorrect C. (1) Correct (2) Correct D. (1) Correct (2) Incorrect QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 215005 APRM / LPRM A2.06 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 or operations: Recirculation flow channels upscale Importance Rating: 3.4 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A Monticello NRC Written Exam Page: 110 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 37 K/A Match: The applicant is given indications of a recirculation flow channel upscale failure and asked to evaluate the effect on interrelated systems. The condition is bypassed by the ARP which will remove the rod block.

SRO Justification: N/A Technical

References:

C.6-005-A-30, C.6-005-A-4, C.6-005-A-22 B.05.01.02-01, B.05.05-02, NH-36243.

Proposed references to be provided: None Learning Objective: M-8107L-066 LO 5a, 7, 8 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 111 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 38 Points: 1.0 The plant is operating in Mode 1 when a loss of #11 Recirc pump occurs. ALL APRMs are adjusted for single loop operation (SLO). Given the following:

• Operating BELOW the MELLA+ region

• APRM 2/4 ODA indicates Total Recirc flow is 41%

Which one on the following approximates the current STP SCRAM Setpoint?

A. 72.2%

B. 78.2%

C. 81.6%

D. 90.2%

Monticello NRC Written Exam Page: 112 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 38 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A. Incorrect: This number represents the APRM STP UPSCALE Alarm/Rod Block single loop operation (SLO) flow-biased setpoint using the equation 0.55W+49.6% for 41% flow.

(55*0.41)+49.6=72.15%

B. Correct: With one recirculation pump running the single loop equation (SLO) must be used to determine the STP Flow Biased setpoints. For the UPSCALE TRIP/SCRAM setpoint this equation is 0.55W+55.6%. W is equal to percent total recirc flow which in this case is 41%.

(55*0.41)+55.6 = 78.15%.

C. Incorrect: This number represents the OLD APRM STP UPSCALE TRIP/SCRAM SLO flow biased setpoint. According to the RPS (1) Diagram (Big Notes) dated 03/20/2012 shown in M-8107L-072 (Reactor Protection System), Rev. 16, the old equation was 0.66(W-7.6) +59.6 (SLO). (66*(0.41-0.076))+59.6=81.64%

D. Incorrect: This number represents the APRM STP UPSCALE TRIP/SCRAM Two Loop Operation (TLO) flow-biased setpoint using the equation 0.61W+65.2% for 41% flow.

(61*0.41)+65.2=90.21%

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (215005 APRM / LPRM) A1.04 Ability to predict and/or monitor changes in parameters associated with operating the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including: (CFR:

41.5 / 45.5) SCRAM and rod block trip setpoints Importance Rating: 4.1 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants ability to predict how the current conditions will affect the APRM SCRAM setpoints SRO Justification: N/A Monticello NRC Written Exam Page: 113 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 38 Technical

References:

Ops Man ARP 5-A-22, B.05.01.02, B.05.06-02, M-8107L-066 (Power Range Neutron Monitoring System) Rev. 10, M-8107L-072 (Reactor Protection System), Rev. 16 Proposed references to be provided: None Learning Objective: 7 from M-8107L-066 (Power Range Neutron Monitoring System) Rev. 10 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Question 1 of 2010 MNGP ILT NRC Written Exam, and Question 34 of 2013 MNGP ILT NRC Written Exam Comments:

Monticello NRC Written Exam Page: 114 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 39 Points: 1.0 The plant was at rated conditions with HPCI out of service when a Group I Isolation occurred.

Given the following:

• RCIC received an auto initiation signal on RPV Low-Low Level 5 minutes later:

• The RCIC pump flow signal input to the flow controller failed HIGH Assuming that no operator action has been, or will be taken, complete the following statement describing the RCIC system response.

The RCIC turbine speed will....

A. remain relatively constant B. Lower to approximately 2000 rpm C. Rise to approximately 4500 rpm and will subsequently trip on high RPV water level D. Rise until the mechanical overspeed trips the turbine Monticello NRC Written Exam Page: 115 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 39 ANSWER EXPLANATION ANSWER B DISTRACTOR ANALYSIS In automatic flow control mode, the RCIC system flow controller compares the pump flow with the controller-setpoint and generates a signal proportional to the difference. Controller output of 4 mA to 20 mA corresponds to turbine speeds of 2000 rpm to 4500 rpm, respectively. In this condition, the RCIC turbine speed would decrease due to the high flow signal and continue to operate at 2000 rpm. The mechanical overspeed would not be reached (5625 RPM). High water level would not be reached because the RCIC turbine is operating at minimum speed A. Incorrect, plausible if the applicant believes the controller is normally in manual B. Correct C. Incorrect, this would occur if the signal failed low D. Incorrect, rpm is limited to 4500 by the governor, which did not fail.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 217000 RCIC A3.04 Ability to monitor automatic operations of the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) including: System flow Importance Rating: 3.6 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: The applicant is asked to evaluate how the RCIC system responds to a failure of the flow signal to the controller.

SRO Justification: N/A Monticello NRC Written Exam Page: 116 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 39 Technical

References:

B.02.03-01, B.02.03-02 Proposed references None to be provided:

Learning Objective: M-8107L-003, Rev. 18, learning objective 7.

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: ILT Bank - 2013 NRC Exam, question 38 Comments:

Monticello NRC Written Exam Page: 117 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 40 Points: 1.0 The plant is at 80% power with RCIC Full Flow Testing being conducted and the following line-up:

• MO-2075 (RCIC STEAM LINE ISOLATION INBOARD) OPEN

• MO-2076 (RCIC STEAM LINE ISOLATION OUTBOARD) OPEN

• MO-2106 (RCIC PUMP DISCHARGE OUTBOARD) OPEN

• MO-2107 (RCIC PUMP DISCHARGE INBOARD) CLOSED

• MO-2110 (RCIC TEST FLOW ISOLATION) OPEN

• MO-3502 (RCIC TEST RETURN FLOW) OPEN During the RCIC Full Flow test, a loss of the D-31 (Division I 125/250 VDC Distribution Panel) occurs shortly followed by a loss of Feedwater lowering RPV water level to -47.

Which of the following will occur and what action, if any, can be taken to restore RPV water level with RCIC?

A. RCIC will automatically realign for RPV injection, NO action is required.

B. RCIC will remain in the full-flow test lineup, reposition RCIC valves at C-04 for RPV injection.

C. RCIC will remain in the full-flow test lineup, reposition RCIC valves locally for RPV injection.

D. RCIC will trip as a direct result of the loss of DC power, reposition RCIC valves locally for a shutdown lineup.

Monticello NRC Written Exam Page: 118 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 40 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A, B, and D are incorrect: plausible but incorrect interpretations of the RCIC system with loss of DC power: it is plausible that the applicant doesnt know (1) D-31 Division I 125/250 VDC Distribution Panel powers RCIC; (2); RCIC will NOT auto initiate on loss of FW (3) that the DC-powered valves remain as is due to loss of DC; or (4) what the appropriate next step to take is in C.4-B.09.09.A (LOSS OF A 250 VDC BUS).

C is correct: Abnormal Procedure C.4-B.09.09.A shall be entered when D31 is lost and since RCIC is supplied by MCC D311 via D31, the DC valves remain as is in the full flow test line-up due to loss of power.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (217000 RCIC) A2.05 Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) D.C. power loss Importance Rating: 3.3 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants ability to predict the impact a loss of D.C. power will have on the RCIC system and based on those predictions use procedures to mitigate the consequences of the loss of D.C. power SRO Justification: N/A Technical

References:

Ops Man B.02.03-01 and -05, C.4-B.09.09.A, NH-36251, NH-36252, M-8107L-003 (RCIC) Rev. 18, M-8107L-041 (Station Battery Systems) Rev. 14 Monticello NRC Written Exam Page: 119 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 40 Proposed references to be provided: None Learning Objective: 8.g from M-8107L-003 (RCIC) Rev. 18 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: N/A Comments: Although the base of this question comes from Question 69 in the Exam BANK, it has been significantly modified.

Monticello NRC Written Exam Page: 120 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 41 Points: 1.0 The plant was at 100% power when an LOCA occurred with the following timeline of indications:

TIME Drywell pressure Reactor Level 02:00:00 0.5 psig 30 inches 02:10:00 1.9 psig -20 inches 02:20:00 3.5 psig -60 inches 02:30:00 5.5 psig -80 inches 02:40:00 6.5 psig -50 inches 02:50:00 7.0 psig -40 inches Complete the statements below?

The MINIMUM operator action(s) required to reset and close the ADS SRVs is to take __(1)__ to INHIBIT and then back to AUTO; Given the above time line, the EARLIEST this can occur without reinitiating the ADS initiation logic is __(2)__.

A. (1) EITHER the A or B INHIBIT switch (2) 02:40:00 B. (1) EITHER the A or B INHIBIT switch (2) 02:50:00 C. (1) BOTH the A and B INHIBIT switches (2) 02:40:00 D. (1) BOTH the A and B INHIBIT switches (2) 02:50:00 Monticello NRC Written Exam Page: 121 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 41 ANSWER EXPLANATION Answer: D DISTRACTOR ANALYSIS A. (1) Incorrect, Either channel is capable of actuating automatic blowdown, but both must be reset to close the ADS SRVs (2) Incorrect, the low-low level signal is still present so the ADS initiation timer would start B. (1) Incorrect, Either channel is capable of actuating automatic blowdown, but both must be reset to close the ADS SRVs (2) Correct, the low-low level signal would be cleared C. (1) Correct, Both channels must be reset for the ADS SRVs to close (2) Incorrect, the low-low level signal is still present so the ADS initiation timer would start D. (1) Correct, Both channels must be reset for the ADS SRVs to close (2) Correct, the low-low level signal would be cleared QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 218000 ADS A4.03 Ability to manually operate and/or monitor in the control room: ADS logic reset Importance Rating: 4.2 10 CFR Part 55: 55.41.7 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine the minimum action required to reset the ADS logic which results in the closure of ADS SRVs that would be aligned to auto, and to determine that the initiating signal of low-low reactor water level must be cleared.

SRO Justification: N/A Monticello NRC Written Exam Page: 122 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 41 Technical

References:

OPS MAN B.03.03-01 OPS MAN B.03.03-05 Proposed references to be provided: None Learning Objective: M-8107L-005 LO 7 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History: New Comments:

Monticello NRC Written Exam Page: 123 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 42 Points: 1.0 Which of the following PCIS Group Isolations have MANUAL INITIATION capability through the use of pushbutton(s) in the Control Room?

A. Group 3 ONLY B. Group 4 ONLY C. Groups 1 and 4 ONLY D. Groups 1, 3 and 4 Monticello NRC Written Exam Page: 124 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 42 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS The HPCI System has a C-03 pushbutton that allows for the manual insertion of a Group 4 isolation as long as an initiation signal is present.

A and C are incorrect: Plausible for confusion between HPCI and RWCU (Group 3). RWCU has a spring return switch for resetting the Group 3 but not a manual isolation pushbutton.

D is incorrect: The MSIVs (Group 1) have pushbuttons to reset power to the isolation solenoids but not a manual initiation pushbutton.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (223002 PCIS/Nuclear Steam Supply Shutoff) 2.1.27 Knowledge of system purpose and/or function Importance Rating: 3.9 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the PCIS function SRO Justification: N/A Technical

References:

Ops Man B.03.02-03 Proposed references to be provided: None Monticello NRC Written Exam Page: 125 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 42 Learning Objective: M-8107L-002 (HPCI)

Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank _

Bank X Question History: 2010 MNGP NRC Exam #43 Comments:

Monticello NRC Written Exam Page: 126 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 43 Points: 1.0 Given the following conditions:

• A loss of Feedwater has occurred.

• RPV water level lowered to -49 inches and is recovering with HPCI and RCIC.

• Reactor pressure is 900 psig.

• The CRS has directed you to verify expected initiations and isolations.

Which ONE of the following valves is out of position for the given plant conditions AND should be repositioned?

A. MO-2013 (RHR DIV 2 LPCI INJ OTBD) is OPEN.

B. MO-1753 (11 CS INJECTION INBOARD) is CLOSED.

C. MO-2107 (RCIC PUMP DISCHARGE INBOARD) is OPEN.

D. MO-2373 (MAIN STEAM LINE DRAIN INBOARD) is OPEN.

Monticello NRC Written Exam Page: 127 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 43 ANSWER EXPLANATION Answer D DISTRACTOR ANALYSIS A. Incorrect, M0-2013 is normally open, remains open because LPCI Loop Select chose "B" Loop to inject into.

B. Incorrect, MO-1753 is normally closed and would not open until reactor pressure was <420 psig with an ECCS signal present C. Incorrect, MO-2107 is open because it auto opened on Low Low RPV water level.

D. Correct, MO-2373 should have closed on the Group I isolation signal of Low Low RPV Water Level.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 223002 PCIS/Nuclear Steam Supply Shutoff K1.01 Knowledge of the physical connections and/or cause effect relationships between PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF and the following: Main steam system Importance Rating: 3.8 10 CFR Part 55: 55.41.2 to 41.9 10 CFR 55.43.b  : N/A K/A Match: KA is matched by having the applicant determine that plant conditions have warranted a group 1 isolation, and the effect of this is that auto closure of the main steam drain isolation valves should have occurred.

SRO Justification: N/A Technical

References:

Plant Protection System, Ops Man B.05.06-01 Core Spray Cooling System, Ops Man B.03.01-01 Residual Heat Removal System, Ops Man B.03.04-02 Reactor Core Isolation Cooling, Ops Man B.02.03-02 Monticello NRC Written Exam Page: 128 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 43 Proposed references None to be provided:

Learning Objective: M8107L-070 24 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: MNGP Bank question 1198 Comments:

Monticello NRC Written Exam Page: 129 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 44 Points: 1.0 Complete the following statement:

SRVs __(1)__ are assigned to the Automatic Depressurization System (ADS) Logic Channel A and SRVs __(2)__ are assigned to ADS Logic Channel B.

A. (1) A, C, and D (2) E, G, and H B. (1) E, G, and H (2) A, B, and C C. (1) A, C, and D (2) A, C, and D D. (1) E, G, and H (2) E, G, and H Monticello NRC Written Exam Page: 130 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 44 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A, B and D are incorrect: Three SRVs are required for the ADS function. Choices are plausible if the candidate believes that each logic channel operates different valves and/or the candidate confuses the ADS assigned valves with the Low-Low-Set assigned valves.

C is correct: There are eight (8) safety/relief valves (SRV) located on the main steam lines within the drywell between the reactor vessel and the first main steam isolation valves (MSIV).. Three of the SRVs (RV-2-71A, C & D) are assigned to the Automatic Depressurization System (ADS) and the valves are operated by either or both logic channels.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (239002 SRVs) K1.08 Knowledge of the physical connections and/or cause effect relationships between RELIEF/SAFETY VALVES and the following: (CFR: 41.2 to 41.9 / 45.7 to 45.8) Automatic depressurization system Importance Rating: 4.0*

10 CFR Part 55: 41.2 to 41.9 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of which SRVs are assigned to the Automatic Depressurization System function.

SRO Justification: N/A Technical

References:

Ops Man B.03.03-01, NH-36241-1, M-8107L-025 (Reactor Pressure Relief)

Rev. 23 Proposed references to be provided: None Monticello NRC Written Exam Page: 131 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 44 Learning Objective: 5 and 6 from M-8107L-025 (Reactor Pressure Relief) Rev. 23 Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History: N/A Comments:

Monticello NRC Written Exam Page: 132 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 45 Points: 1.0 A plant startup is in progress following a refuel outage. Given the following:

• Both Main Feedwater Regulating Valves are in AUTO

• Master Controller is in AUTO with the setpoint at 38 inches

• DFCS is in 3-Element Predict what actual RPV water level (inside the shroud) will be when reactor power is raised to 100%?

A. 28 inches B. 31 inches C. 35 inches D. 38 inches Monticello NRC Written Exam Page: 133 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM ANSWER EXPLANATION Answer A DISTRACTOR ANALYSIS The 10" depression in actual reactor water level which is obtained over a range of 10% to 100% reactor power results, in part, from the steam flow pressure drop which exists across the steam dryer. As reactor steam flow increases from 10% to 100%, the pressure drop across the steam dryer increases, resulting in a seven inch depression in water level at full power. Since water level is monitored outside of the separator plenum area (see Figure 2), this seven inch drop is not detected by the level monitoring instrumentation. The remainder of the 10 inch drop (3 inches) is obtained by means of the steam flow programmer (see Figure 1).

As steam flow changes from 10% to 100%, the programmer reduces the level setpoint of the main feedwater controller by three inches. This three inch change in level is indicated on all reactor level indicators since it occurs outside of the separator plenum area as well as inside.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 259002 Reactor Water Level Control System A3.05 Ability to monitor automatic operations of the Reactor Water Level Control System including: Changes in reactor power Importance Rating: 3.4 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A K/A Match: KA is matched by first having the applicant evaluate the change in programmed level as a function of reactor power.

Monticello NRC Written Exam Page: 134 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 45 SRO Justification: N/A Technical

References:

Ops Man B.05.07-02 Proposed references None to be provided:

Learning Objective: M-8107L-046 LO 6, 7 Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank _

Bank X Question History: Monticello Exam Bank Q873 (M8107L-046 4)

Comments:

Monticello NRC Written Exam Page: 135 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 46 Points: 1.0 The plant was at full power with SBGT Train B isolated when a DBA LOCA occurred.

All automatic actuations/isolations occurred as expected; EXCEPT SBGT Train A failed to start and cannot be started.

Without further operator action, the Secondary Containment differential pressure is expected to

__(1)__ and any release from the Secondary Containment to the environment will be __(2)__.

(1) (2)

A. lower but remain negative filtered and elevated B. lower and then go negative filtered and elevated C. lower but remain positive unfiltered but elevated D. lower and then go positive unfiltered and at ground level Monticello NRC Written Exam Page: 136 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 46 ANSWER EXPLANATION Answer: D DISTRACTOR ANALYSIS D is correct: In the event of an accident, which requires that SCT integrity be established, the SBGT system SHALL establish a sub atmospheric pressure in SCT and provide a filtered SCT exhaust flow path to the Off-Gas Stack; however SBGT is INOP so unfiltered air will remain in the SCT and a negative pressure CANNOT be maintained (differential pressure between the RB and atmosphere will increase as the pressure builds within the RB). As a result, the SCT is not expected to minimize ground level escape of airborne radioactivity.

A, B, and C are incorrect: plausible interpretations of what effects a loss of the SBGT will have: with both trains of the SBGT system INOP, the SCT differential pressure will decrease and go positive.

With normal ventilation trains shutdown and/or isolated air is expected to leak from the SCT at ground level.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (261000 SGTS) K3.01 Knowledge of the effect that a loss or malfunction of the STANDBY GAS TREATMENT SYSTEM will have on following: (CFR: 41.7 /45.6)

Secondary containment and environment differential pressure Importance Rating: 3.3 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the effect a loss of the Standby Gas Treatment System will have on the Secondary Containment and environment differential pressurize (pressure across the RB)

SRO Justification: N/A Monticello NRC Written Exam Page: 137 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 46 Technical

References:

Ops Man B.04.02-01, M-8107L-008 (Standby Gas Treatment and Secondary Containment Systems) Rev. 19 Proposed references to be provided: None Learning Objective: 8.a.vii from M-8107L-008 (Standby Gas Treatment and Secondary Containment Systems) Rev. 19 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History: N/A Comments:

Monticello NRC Written Exam Page: 138 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 47 Points: 1.0 The plant is operating at 100% power when a sudden pressure fault condition occurred in the generator main transformer (Sudden Pressure Relay, SPR-63, actuates)

Which of the following completes the statements below?

This condition DIRECTLY actuates the __(1)__.

The field breaker trips open __(2)__.

A. (1) turbine lockout relay (286/T)

(2) immediately B. (1) turbine lockout relay (286/T)

(2) once 8N7 and 8N8 are sensed open C. (1) generator lockout relay (286/G)

(2) immediately D. (1) generator lockout relay (286/G)

(2) once 8N7 and 8N8 are sensed open Monticello NRC Written Exam Page: 139 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 47 ANSWER EXPLANATION Answer C DISTRACTOR ANALYSIS A. (1) Incorrect, the turbine lockout relay 286/T is actuated by the generator lockout relay 286/G (2) Correct, generator lockout relay 286/G trips and locks out the field breaker B. (1) Incorrect, the turbine lockout relay 286/T is actuated by the generator lockout relay 286/G when (2) Incorrect, generator lockout relay 286/G trips and locks out the field breaker. Would be correct for any turbine trip not caused by a generator lockout.

C. (1) Correct, SPR-63; Generator transformer sudden pressure will operate for fault in the generator transformer and operation of this relay will cause the generator lockout relay 286/G to trip (2) Correct, generator lockout relay 286/G trips and locks out the field breaker D. (1) Correct, SPR-63; Generator transformer sudden pressure will operate for fault in the generator transformer and operation of this relay will cause the generator lockout relay 286/G to trip (2) Incorrect, generator lockout relay 286/G trips and locks out the field breaker. Would be correct for any turbine trip not caused by a generator lockout.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 262001 AC Electrical Distribution K4.02 Knowledge of A.C. ELECTRICAL DISTRIBUTION design feature(s) and/or interlocks which provide for the following: Circuit breaker automatic trips Importance Rating: 2.9 10 CFR Part 55: 55.41.7 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine that a sudden overpressure condition in the main transformer will result in generator lockout trip and that the field breaker is immediately opened on that trip condition.

Monticello NRC Written Exam Page: 140 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 47 SRO Justification: N/A Technical

References:

Ops Man B.09.02-02, Generation Proposed references to be provided: None Learning Objective: M-8107L-036 LO7 Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank Question History:

Comments:

Monticello NRC Written Exam Page: 141 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 48 Points: 1.0 The plant is at rated conditions when the following occurs:

• 8-A-29 (DIV II INVERTER Y-81 TROUBLE) is received

• The FUSE BLOWN indicator is illuminated on UPS Inverter Panel Y-81

• The Division II 120 VAC UPS System has responded as expected Which of the following describes how the above conditions will impact components, if at all, associated with the Primary Containment Isolation System (PCIS)?

A. PCIS will be UNAFFECTED.

B. A partial RWCU Group 3 Isolation will occur.

C. The SBGT System will start and Secondary Containment will isolate.

D. The High Temperature Isolation of the RWCU System will be blocked.

Monticello NRC Written Exam Page: 142 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 48 ANSWER EXPLANATION Answer: A DISTRACTOR ANALYSIS A. Correct: The inverters will transfer to the alternate source automatically and power to the distribution panels will not be lost. Additionally, placing the MBS in Bypass is Make-Before-Break. Since power is not lost to the distribution panels, components associated with the Containment Isolation System will NOT be affected.

B. Incorrect: The inverters will transfer to the alternate source automatically and power to the distribution panels will not be lost; if power were lost to Y-80, a partial RWCU system isolation would occur.

C. Incorrect: The inverters will transfer to the alternate source automatically and power to the distribution panels will not be lost; if power were to lost to Y-80, the SBGT System would start and the Secondary Containment would isolate.

D. Incorrect: There is no loss of sync and the inverters will transfer to the alternate source automatically and power to the distribution panels will not be lost; if power were to be lost to Y-30, automatic isolation of the RWCU System on High Filter/Demineralizer Inlet Temperature will be blocked due to loss of power to logic relay.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: K3.10 Knowledge of the effect that a loss or malfunction of the UNINTERRUPTABLE POWER SUPPLY (A.C./D.C.) will have on following: (CFR:

41.7 / 45.4) Containment isolation: Plant-Specific Importance Rating: 2.7 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the effects that a loss or malfunction of UPS (Y-81) will have on containment isolation Monticello NRC Written Exam Page: 143 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 48 SRO Justification: N/A Technical

References:

Ops Man B B.09.13-01 and -06, B.02.04-05, 8-A-29, M-8107L-063 (Uninterruptable Power Supply and Instrument AC) Rev 12, M-8107L-070 (Primary Containment Isolation System) Rev. 12 Proposed references to be provided: None Learning Objective: 7.b and 8.a from M-8107L-063 (Uninterruptable Power Supply and Instrument AC) Rev 12 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: N/A Comments: Question has been significantly modified from Question 1065 from Exam BANK. One of the distractors may need some work.

Monticello NRC Written Exam Page: 144 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 49 Points: 1.0 The plant is at rated conditions.

If ALL AC power is lost to LC-107 and LC-108; Complete the following statements:

D71 (250 VDC DISTRIBUTION PANEL) is powered from __(1)__.

With no operator action, Y91 (480 VAC UPS) will supply all Y94 (480 VAC DISTRIBUTION PANEL) loads for __(2)__ minutes.

A. (1) Battery #17 ONLY (2) 30 B. (1) Battery #17 ONLY (2) 60 C. (1) Battery #17 AND Y-91 (2) 30 D. (1) Battery #17 AND Y-91 (2) 60 Monticello NRC Written Exam Page: 145 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 49 ANSWER EXPLANATION Answer A DISTRACTOR ANALYSIS A. (1) Correct (2) Correct B. (1) Correct (2) Incorrect, on a loss of AC power to Y-91, the process computer panel supplies are automatically shed after 30 minutes to extend the availability of 250 VDC Battery 17.

C. (1) Incorrect, The AC source that supplies the battery is from LC-108 to Y-91, LC-107 it the alternate supply to Y-91s output to Y-94 (2) Correct D. (1) Incorrect, The AC source that supplies the battery is from LC-108 to Y-91, LC-107 it the alternate supply to Y-91s output to Y-94 (2) Incorrect, on a loss of AC power to Y-91, the process computer panel supplies are automatically shed after 30 minutes to extend the availability of 250 VDC Battery 17.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 263000 DC Electrical Distribution K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the D.C. ELECTRICAL DISTRIBUTION: A.C. electrical distribution Importance Rating: 3.2 10 CFR Part 55: 55.41.7 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant determine that the normal AC power supply to Y-91 is lost. The charger for battery 17 and its associated bus have lost power from the output of Y-91 will only have battery power. Subsequently the alternate power supply to Y-94 is lost and the effects of this are there is not power to the plant computer since greater than 30 minutes have elapsed since the normal power supply to Y-91 was lost.

Monticello NRC Written Exam Page: 146 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 49 SRO Justification: N/A Technical

References:

Ops Man B.09.13-02 Ops Man B.09.13-01 Proposed references to be provided: None Learning Objective: M-8107L-063 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 147 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 50 Points: 1.0 The plant is at rated conditions with the 1AR Transformer ISOLATED for modifications. A loss of normal off-site power (LONOP) then occurs coincident with an ECCS initiation signal. Complete the following statement for the above conditions.

Nominally, #13 and #14 RHR Pumps will auto start _________ after the LONOP.

A. 10 seconds B. 15 seconds C. 20 seconds D. 25 seconds Monticello NRC Written Exam Page: 148 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 50 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A. incorrect: Plausible time for the EDG start time B. incorrect: Plausible start time for 11 and 12 RHR pumps C. correct: If normal auxiliary electrical power is not available at the time of an ECCS automatic initiation, a 10 second delay is added to allow the two EDG to start, come up to speed and for their respective breakers to CLOSE. This will be followed by a 10 second delay before starting 13 & 14 RHR Pumps. A total of 20 seconds will lapse before 13 and 14 RHR pumps start.

According to B.03.04-02, Electrical Power Interlocks:

If normal auxiliary electrical power is available at the time of an ECCS automatic initiation, the starting of the RHR and Core Spray pumps will be delayed to limit the maximum motor starting load. The time delays employed are as follows:

Time Essential Bus 15 Essential Bus 16 5 seconds 11 RHR Pump 12 RHR Pump 10 seconds 13 RHR Pump 14 RHR Pump 15 seconds 11 Core Spray Pump 12 Core Spray Pump D. incorrect: Plausible start times for Core Spray pumps based on the power loss.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (264000 EDGs) K5.06 Knowledge of the operational implications of the following concepts as they apply to EMERGENCY GENERATORS (DIESEL/JET): (CFR:

41.5 / 45.3) Load sequencing Importance Rating: 3.4 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A Monticello NRC Written Exam Page: 149 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 50 K/A Match: K/A is matched by testing the applicants knowledge of the load sequencing following Emergency Diesel Generator initiation SRO Justification: N/A Technical

References:

Ops Man B.03.04-02, B.09.08-01, M-8107L-042 (Diesel Generator System)

Rev. 19 Proposed references to be provided: None Learning Objective: 8 from M-8107L-042 (Diesel Generator System), Rev. 19 Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank _

Bank X Question History: Question 49 of the 2013 MNGP ILT NRC Written Exam Comments:

Monticello NRC Written Exam Page: 150 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 51 Points: 1.0 Given the following plant conditions

• Instrument Air system pressure is 102 psig with 15 A/C in Lead

• 6-B-20 (15 AIR DRYER S-121 TROUBLE) is in alarm

• S-121 (15 INST & SERVICE AIR DRYER) drying is in FIXED Mode and dew point is 10°F Complete the statements below?

S-121 will automatically cycle between chambers based on __(1)__.

The 15 air compressor/dryer train should __(2)__.

A. (1) the level of moisture in the chamber (2) be removed from lead (priority 1) operation B. (1) the level of moisture in the chamber (2) remain in lead (priority 1) operation until dew point is < 0°F C. (1) a fixed time interval (2) be removed from lead (priority 1) operation D. (1) a fixed time interval (2) remain in lead (priority 1) operation until dew point is < 0°F Monticello NRC Written Exam Page: 151 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 51 ANSWER EXPLANATION Answer C DISTRACTOR ANALYSIS A. (1) Incorrect, AMLOC Mode [NORMAL] dryer will automatically cycle between chambers based on the level of moisture in the chamber as sensed on the AMLOC (2) Correct, if the dew point from one of the dryers is 0°F, the dryer should be removed from lead (priority 1) operation B. (1) Incorrect, AMLOC Mode [NORMAL] dryer will automatically cycle between chambers based on the level of moisture in the chamber as sensed on the AMLOC (2) Incorrect, if the dew point from one of the dryers is 0°F, the dryer should be removed from lead (priority 1) operation C. (1) Correct, The dryer will still switch back and forth between chambers on a fixed timer, and it will purge and recharge each time (2) Correct, if the dew point from one of the dryers is 0°F, the dryer should be removed from lead (priority 1) operation D. (1) Correct, The dryer will still switch back and forth between chambers on a fixed timer, and it will purge and recharge each time (2) Incorrect, if the dew point from one of the dryers is 0°F, the dryer should be removed from lead (priority 1) operation QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 300000 Instrument Air A2.01 Ability to (a) predict the impacts of the following on the INSTRUMENT AIR SYSTEM and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal operation: Air dryer and filter malfunctions Importance Rating: 2.9 10 CFR Part 55: 55.41.5 10 CFR 55.43.b: N/A K/A Match: The applicant is given indications of an air dryer malfunction and asked what affects that may have on the regeneration cycle and if the dew point of the dryer outlet is acceptable for continued lead operation.

Monticello NRC Written Exam Page: 152 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 51 SRO Justification: N/A Technical

References:

Ops Man B.08.04.01-05 Ops Man B.08.04.01-01 Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 153 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 52 Points: 1.0 The plant is operating in MODE 1. Due to fouling of the in-service RBCCW Heat Exchanger tubes, the Service Water inlet and outlet valves to the heat exchanger are now full open and Service Water flow has gradually lowered.

If the declining trend in Service Water flow through the RBCCW Heat Exchangers (HX) continues, which one of the following describes how the RWCU system will be affected?

A. RWCU Pump Room temperatures will rise, eventually causing RWCU to isolate and the RWCU pumps to trip.

B. The RWCU System temperature at discharge of the RWCU pumps will rise, eventually causing RWCU to isolate and the RWCU pumps to trip.

C. Upon receipt of a High RBCCW HX Outlet temperature signal, the RWCU Filter Demineralizers will automatically go into hold and RWCU system flow will bypass the Filter Demineralizers using MO-2400 (Filter/Demineralizer Bypass).

D. RBCCW System Flow through the RWCU NRHX will automatically increase in response to the rising RWCU system temperature.

Monticello NRC Written Exam Page: 154 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 52 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A. Incorrect: Neither RBCCW nor Service Water supply the area room coolers.

B. Correct: If RBCCW temperature rises (system is nominally maintained <95F), the RWCU water temperature will also rise and Inlet Filter Demineralizer water temperature will eventually be > 140°F, causing a HI-HI Demin water temperature condition to alarm (4-B-

31) and stopping RWCU water flow by closing valves MO-2399 (per B.02.02-01).

C. Incorrect: There are no automatic control features associated with MO-2400.

D. Incorrect: There are no automatic temperature/flow control valves on the RBCCW outlet of the RWCU NRHS.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (400000 Component Cooling Water) A1.02 Ability to predict and / or monitor changes in parameters associated with operating the CCWS controls including:

(CFR: 41.5 / 45.5) CCW temperature Importance Rating: 2.8 10 CFR Part 55: 41.5 10 CFR 55.43.b  : N/A K/A Match: K/A is matched by testing the applicants ability to monitor the CCW temperature and predict the changes an altering temperature will have on the operating CCWS SRO Justification: N/A Technical

References:

Ops Man B.02.05.01, B.02.02-01, 4-B-31, M-8107L-026 (RBCCW) Rev. 14, M-8107L-030 (Reactor Water Cleanup System) Rev. 18 Proposed references to be provided: None Monticello NRC Written Exam Page: 155 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 52 Learning Objective: 5.g from M-8107L-026 (RBCCW) Rev. 14 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Question 687 from the Exam BANK Comments: From Question 687 of the Exam BANK, the distractors have been slightly altered and Distractor D has been replaced.

Monticello NRC Written Exam Page: 156 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 53 Points: 1.0 The plant is at rated conditions with 11 RBCCW pump in service & 12 RBCCW pump in Standby.

Which of the following describes the RBCCW System response to a Loss of ALL Offsite Power?

A. The 11 and 12 RBCCW pumps will automatically start when the associated EDG restores power to the respective 4.16 KV BUS.

B. The EBT Load Shed ONLY must be reset to manually restart a RBCCW pump.

C. The EBT & ECCS Load Shed must be reset to manually restart a RBCCW pump.

D. The 11 and 12 RBCCW pumps can be manually restarted when the associated EDG restores power to the respective 4.16 KV BUS without resetting any Load Shed circuits.

Monticello NRC Written Exam Page: 157 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 53 ANSWER EXPLANATION Answer D DISTRACTOR ANALYSIS A. Incorrect, neither RBCCW will auto start upon power restoration. The standby pump will not auto start since the standby logic drops out during the LONOP.

B. Incorrect; the Essential Bus Transfer load shed does not load shed RBCCW.

C. Incorrect; A loss of all offsite power would only result in an Essential Bus Transfer and not an ECCS Load Shed. If there was an ECCS load shed and an EBT, the RBCCW pump would be load shed.

D. Correct - once power is restored, 11 RBCCW can be manually started. The standby pump will not auto start since the standby logic drops out during the LONOP.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 400000 Component Cooling Water K6.05 Knowledge of the effect that a loss or malfunction of the following will have on the CCWS: Motors Importance Rating: 2.8 10 CFR Part 55: 55.41.7 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant know the effect that a loss of power and subsequent restoration has on the RBCCW pump system.

SRO Justification: N/A Monticello NRC Written Exam Page: 158 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 53 Technical

References:

Ops Man C.4-B.02.05.A Proposed references None to be provided:

Learning Objective: M-8107L-026 LO 4, 7 Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank X Bank _

Question History: MNGP Bank 584 Comments:

Monticello NRC Written Exam Page: 159 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 54 Points: 1.0 Which of the following causes the Scram Discharge Volume (SDV) Vent and Drain valves to close during a reactor scram?

A. Energizing SDV pilot solenoid valves from the 125 VDC batteries.

B. De-energizing SDV pilot solenoid valves from the 125 VDC batteries.

C. De-energizing SDV Pilot solenoid valves from the Reactor Protection System bus.

D. Energizing Backup Scram solenoid valves from the Reactor Protection System bus.

Monticello NRC Written Exam Page: 160 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 54 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A, B, and D are incorrect but plausible interpretations of the power supply and functioning of the SDV vent and drain valve solenoids during a scram: 125 VDC batteries power the backup scram pilot valves and are energized during scram.

C is correct: During normal operation the pilot valves for the SDV vent and drain valves are energized. Upon the initiation of a Reactor scram, both Reactor Protection System logic channels are de-energized and BOTH pilot valves (in both sets) open, venting control air from the vent and drain valve operators and permitting these valves to close.

QUESTION BACKGROUND Tier: 2 Group: 2 K/A: (201001 CRD Hydraulic) K2.04 Knowledge of electrical power supplies to the following: (CFR: 41.7) Scram discharge volume vent and drain valve solenoids Importance Rating: 3.2 10 CFR Part 55: 41.7 10 CFR 55.43.b  : N/A K/A Match: K/A is matched by testing the applicants knowledge of the power supplies to the scram discharge volume vent and drain valve solenoids SRO Justification: N/A Technical

References:

Ops Man B.01.03-01, M-8107L-020 (CRD Hydraulics) Rev. 17 Proposed references to be provided: None Learning Objective: 4 from M-8107L-020 (CRD Hydraulics) Rev. 17 Monticello NRC Written Exam Page: 161 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 54 Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank _

Bank X Question History: Question 379 from the Exam BANK Comments:

Monticello NRC Written Exam Page: 162 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 55 Points: 1.0 A plant startup is in progress at the POAH with IRMs on Range 5.

Subsequently the reactor becomes subcritical without rod motion requiring the OATC to downrange the IRMs to Range 3.

Which of the following is correct IAW C.1 (STARTUP PROCEDURE) for the conditions above?

A. Insert a manual reactor scram.

B. Withdraw control rods back to the POAH.

C. Continue ranging IRMs until power stabilizes.

D. Insert control rods to the subcritical hold point.

Monticello NRC Written Exam Page: 163 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 55 ANSWER EXPLANATION Answer D DISTRACTOR ANALYSIS After reaching the POAH, a subcritical reactor is indicated by multiple IRMs have been down ranged through at least 2 ranges. Ops Man C.1 part III.F (Operator Response to Subcriticality) states:

If subcriticality occurs, Operator action SHALL be taken to prevent an unintentional re-criticality in the event Reactor water temperature decreases.

In this condition sufficient rod notches should be inserted to maintain the Reactor subcritical (i.e. all IRMs on Range 1). Control rods should be inserted until the subcritical hold point noted on Form 2159.

The SRMs SHALL be fully inserted to permit full range neutron flux level monitoring.

A. Incorrect B. Incorrect C. Incorrect D. Correct QUESTION BACKGROUND Tier: 2 Group: 2 K/A: 201003 Control Rod and Drive Mechanism K5.03 Knowledge of the operational implications of the following concepts as they apply to CONTROL ROD AND DRIVE MECHANISM: Reactor power control Importance Rating: 3.3 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A Monticello NRC Written Exam Page: 164 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 55 K/A Match: KA is matched by placing the applicant in a startup condition with the control rods responsible for reactor power control. The applicant must remember that an indication of a subcritical reactor when in the intermediate range is multiple IRMs have been down ranged through at least 2 ranges. The applicant must then state the requirement to prevent inadvertent recriticality given applicable conditions.

SRO Justification: N/A Technical

References:

OPS MAN C.1 Proposed references to be provided: None Learning Objective: M-8107L-021 LO 8 Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 165 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 56 Points: 1.0 A Reactor startup is in progress and control rod withdrawal has commenced. Control rod 26-27 is selected and the appropriate ROD SELECT lamps turn ON. A transient occurs resulting in the following conditions:

• 5-A-35 (RWM ROD BLOCK)

• 5-B-52 (TORUS WTR HI TEMP SPOTMOS TROUBLE)

• Control rod 26-27 will NOT withdraw Which of the following would cause the above conditions?

A. Loss of power to Panel Y10 B. Loss of power to Panel Y70 C. RPIS Master Clock malfunction D. Steam Flow sensors failing HIGH Monticello NRC Written Exam Page: 166 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 56 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A. Incorrect: Panel Y10 supplies the RPIS, not RWM. Per AOP C.4-B.09.13.A, Loss of Y10 will result in the inability to select control rods and indicating white select light on the full core display will be lost.

B. Correct: Per AOP C.4-B.09.13.F, Loss of Y70 will result in the following: unable to move control rods (RWM inputs to RMCS so insert and withdrawal rod blocks will result); the following annunciators will ALARM: 5-A-35 (RWM ROD BLOCK) and 5-B-52 (TORUS WTR HI TEMP SPOTMOS TROUBLE); and controls to RWM will be lost.

C. Incorrect: A Master Clock malfunction causes RPIS to be inoperative. However, with the RPIS INOP, the control rods would not able to be selected.

D. Incorrect: RWM uses steam flow sensors from the digital feedwater system; however a HIGH reading will NOT cause a rod block (RWM not enforcing with Rx Power >20%);

conversely, a LOSS of steam flow signal would enforce an insert and withdrawal rod block QUESTION BACKGROUND Tier: 2 Group: 2 K/A: (201006 RWM) K6.01 Knowledge of the effect that a loss or malfunction of the following will have on the ROD WORTH MINIMIZER SYSTEM (RWM): (CFR: 41.7 /

45.7) RWM power supply: P-Spec (Not-BWR6)

Importance Rating: 2.8 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the power supply to the Rod Worth Minimizer and the operational implications when that power supply is lost Monticello NRC Written Exam Page: 167 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 56 SRO Justification: N/A Technical

References:

Ops Man B.05.02-03, C.4-B.09.13.F, C.4-B.09.13.A, M-8107L-001 (Rod Worth Minimizer) Rev. 12 Proposed references to be provided: None Learning Objective: 4.e from M-8107L-001 (Rod Worth Minimizer) Rev. 12 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Question 1032 of the Exam BANK Comments: Question has been significantly modified from Question 1032 of the Exam BANK, could be considered a NEW question.

Monticello NRC Written Exam Page: 168 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 57 Points: 1.0 Given the following:

• The plant is at 60% with power ascension in progress

• The #11 Recirc Pump speed control is placed in the RAISE position Complete the statements below?

The maximum Recirc speed is limited in the RAISE direction to __(1)__.

If the Recirc tachometer output signal FAILS HIGH, __(2)__.

A. (1) 2%/second (2) a Scoop Tube Lock will occur B. (1) 2.5%/second (2) a Scoop Tube Lock will occur C. (1) 2%/second (2) recirc pump speed will lower to approximately 50% speed D. (1) 2.5%/second (2) recirc pump speed will lower to approximately 50% speed Monticello NRC Written Exam Page: 169 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 57 ANSWER EXPLANATION Answer A DISTRACTOR ANALYSIS The speed signal from the Flow Control System limits the time rate of change in speed demand to 2.5%/second in the decreasing direction and 2%/second in the increasing direction. Protective logic functions monitor the flow control loop and lock the actuator in position if abnormal conditions exist. This protective circuit will generate a scoop tube lock upon sensing a change in MG set actuator position or MG set speed which was not called for by the control room operator. This failure is MNGP site specific OE (AR 01008012).

A. (1) Correct (2) Correct B. (1) Incorrect (2) Correct C. (1) Correct (2) Incorrect D. (1) Incorrect (2) Incorrect QUESTION BACKGROUND Tier: 2 Group: 2 K/A: 202001 Recirculation A3.05 Ability to monitor automatic operations of the RECIRCULATION SYSTEM including: Pump speed: Plant-Specific Importance Rating: 2.9 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: KA is matched by having the applicant monitor for proper automatic response of the system given an operator initiated demand to raise speed. The system limits the rate speed increase and prevents speed changes is the speed signal from the tachometer is abnormal.

Monticello NRC Written Exam Page: 170 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 57 SRO Justification: N/A Technical

References:

Ops Man B.01.04-01 AR 01008012 Proposed references to be provided: None Learning Objective: M-8107L-047 LO 7, 8 Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank X Bank _

Question History: MNGP bank 915 Comments:

Monticello NRC Written Exam Page: 171 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 58 Points: 1.0 A plant startup is in progress with power at 55% and a control rod selected.

Given the following:

• APRM 2 is bypassed for manual gain adjustment

• Improper data is entered resulting in APRM 2 indication lowering to 22%

Which of the following identifies the expected RBM System response and why?

A. RBM B will NOT enforce rod blocks.

RBM trips are automatically bypassed when APRM 2 dropped below the LPSP.

B. RBM B will continue to enforce rod blocks.

APRM 4 was automatically selected as the reference input to RBM B when APRM 2 was bypassed.

C. RBM B will continue to enforce rod blocks.

RBM B generates an INOP rod block when APRM 2 drops below the LPSP.

D. RBM B will NOT enforce rod blocks.

RBMs will NOT enforce rod blocks when power is >20%.

Monticello NRC Written Exam Page: 172 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 58 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A. Incorrect. This would occur if APRM 2 wasnt bypassed.

B. Correct: When adjusting gains the APRM will be bypassed. When an APRM is bypassed, the RBM will automatically use the backup APRM for its reference. APRM 4 is the backup reference APRM for RBM B.

C. Incorrect: An INOP rod block will not occur.

D. Incorrect: Plausible as the RWM does not enforce rod blocks >20%.

QUESTION BACKGROUND Tier: 2 Group: 2 K/A: (215002 RBM) K1.01 Knowledge of the physical connections and/or cause-effect relationships between ROD BLOCK MONITOR SYSTEM and the following: (CFR:

41.2 to 41.9 / 45.7 to 45.8) APRM: BWR-3, 4, 5 Importance Rating: 2.9 10 CFR Part 55: 41.2 to 41.9 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the connection between the APRMs and the RBMs, and the cause-effects if an APRM malfunctions SRO Justification: N/A Monticello NRC Written Exam Page: 173 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 58 Technical

References:

Ops Man B.05.01.02-01, ARP 5-A-51, M-8107L-066 (Power Range Neutron Monitoring System) Rev. 10 Proposed references to be provided: None Learning Objective: 5 from M-8107L-066 (Power Range Neutron Monitoring System)

Rev. 10 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Question 1124 of the Exam BANK Comments:

Monticello NRC Written Exam Page: 174 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 59 Points: 1.0 The plant was at rated conditions when a reactor scram occurred due to a steam leak inside containment. The following plant conditions now exist:

• RPV Level is (-) 57 inches and slowly rising

• Reactor Pressure is 580 psig and slowly lowering

• Drywell Pressure is 13 psig

• Drywell Temperature is 265 °F The CRS has directed initiation of Drywell Spays using Div 1 RHR.

If the Div 1 CONTAINMENT SPRAY 2/3 CORE HEIGHT BYPASS key-lock switch is unable to be placed in the MANUAL OVERRIDE position, which of the following is correct?

A. Drywell Sprays can NOT be initiated.

B. Drywell Sprays can ONLY be initiated by locally opening the valves.

C. Drywell Sprays CAN still be initiated from C-03; ALL required spray permissives are satisfied.

D. Drywell Sprays CAN still be initiated from C-03 when the CONTAINMENT SPRAY/COOLING LPCI INITIATION BYPASS switch is placed in BYPASS.

Monticello NRC Written Exam Page: 175 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 59 ANSWER EXPLANATION Answer D DISTRACTOR ANALYSIS Containment Spray Isolation Valves (Drywell and Torus) MO-2010, MO-2020, MO-2022, MO-2011, MO-2021 and MO-2023 are all interlocked to be automatically closed when a LPCI automatic initiation signal is received. These valves cannot be opened when a LPCI initiation signal is present except under the following conditions: a) reactor water level is greater than 2/3 core height

[interlock can be bypassed], b) drywell pressure greater than 1 psig, and c) the associated Containment Spray/Cooling LPCI Initiation Bypass switch in BYPASS. The LPCI initiation signal is from PS-10-101A-D and the drywell permissive for core spray originates from PS-10-119A-D.

A. Incorrect, these valves could be opened locally or in the control room with water above 2/3 core height B. Incorrect, these valves could be opened in the control room with water above 2/3 core height C. Incorrect, a valid actuation of LPCI has been received and must be bypassed.

D. Correct, the valves can be opened by momentarily placing the Containment Spray/Cooling LPCI Initiation Bypass switch in Bypass.

QUESTION BACKGROUND Tier: 2 Group: 2 K/A: 226001 RHR/LPCI: CTMT Spray Mode K3.02 Knowledge of the effect that a loss or malfunction of the RHR/LPCI:

CONTAINMENT SPRAY SYSTEM MODE will have on following:

Containment/drywell/suppression chamber temperature Importance Rating: 3.5 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A Monticello NRC Written Exam Page: 176 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 59 K/A Match: KA is matched by having the applicant evaluate the effect of a switch malfunction.

For the given conditions LPCI actuated but RPV water level is above 2/3 core height so Drywell sprays could be initiated.

SRO Justification: N/A Technical

References:

Ops Man C.5-3502 Ops Man B.03.04-01 Ops Man B.03.04-02 Proposed references to be provided: None Learning Objective: M-8107L-023 LO 7 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Browns Ferry Nuclear 2015-301 NRC exam Q 58 Comments:

Monticello NRC Written Exam Page: 177 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 60 Points: 1.0 The plant is in MODE 1 when the following indications are observed:

• Annunciator 3-B-56 (HIGH AREA TEMP STEAM LEAK) alarms

• Temperatures in the Main Steam Chase are rising at approximately 0.5°F/minute Given the above indications, which one of the following:

(1) Correctly characterizes the location of the leak; and (2) Identifies the expected operator response?

A. (1) Outside primary containment (2) Evacuate personnel from the Reactor Building and attempt to identify and isolate the source of the leakage B. (1) Outside primary containment (2) SCRAM the reactor and ensure the MSIVs are closed C. (1) Inside primary containment (2) Attempt to identify and isolate the source of the leakage D. (1) Inside primary containment (2) SCRAM the reactor and ensure the MSIVs are closed Monticello NRC Written Exam Page: 178 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 60 ANSWER EXPLANATION Answer: A DISTRACTOR ANALYSIS A is correct: A main steam line chase temperature of 165°F will sound a HIGH AREA TEMP STEAM LEAK alarm (3-B-56) on Panel C-03 and will help to determine if a steam leak exists outside of Primary Containment. If temperature continues to rise it will give a Group 1 isolation at 195 to 200F; however, the provided trend information indicates that a Group 1 isolation is not expected for approximately 30 minutes. Per ARP 3-B-56, at this point the abnormal procedure C.4-B.02.04.A (Steam Leaks Outside Primary Containment) shall be referred to, which states to evacuate personnel and isolate the leak.

B, C, and D are incorrect but plausible interpretations of the 3-B-56 annunciator received and the plants response to slowly rising main steam line chase temperature. Its plausible that the applicant doesnt know the 3-B-56 alarm applies to the 165F temperature setpoints in the Main Steam areas located outside the Primary Containment. The 3-B-56 alarm is the first indicator of a problem in the steam chase and the setpoint is low enough it could be indicative of a small steam leak (5-10 gpm) in the MSL outside primary containment. The provided trend information indicates that a Group 1 isolation (195-200F) is not expected for approximately 30 minutes. Also, alarms that are more indicative of a LARGE leak/break have not occurred: low pressure and high flow alarms, 5-B-32 (MSL Leakage), and 4-A-11 (RB High Rad). As stated in ARP 3-B-56, AOP C.4-B.02.04.A should be referred to, which states personnel should be evacuated from the affected areas AND the leak should be isolated. C.5-1300 would also be entered but a reactor shutdown is not required until the same parameter in two areas exceed the Max Safe value or it is determined that the leak cannot be isolated.

QUESTION BACKGROUND Tier: 2 Group: 2 K/A: (239001 Main and Reheat Steam) A2.07 Ability to (a) predict the impacts of the following on the MAIN AND REHEAT STEAM SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) Main steam area high temperature or differential temperature high Importance Rating: 3.8 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A Monticello NRC Written Exam Page: 179 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 60 K/A Match: K/A is matched by testing the applicants ability to determine the impacts a rise in the Main Steam area temperature will have on the Main and Reheat Steam System, and what procedural steps to take to mitigate the consequences of this rising temperature SRO Justification: N/A Technical

References:

Ops Man B.02.04-01, ARP 3-B-56, Abnormal Procedure C.4-B.02.04.A, ARP 5-A-17/18, Abnormal Procedure C.4-B.04.01.A, and Emergency Procedure C.5-1300, M-8107L-007 (Main Steam) Rev. 20, M-8107L-070 (Primary Containment Isolation System) Rev. 12 Proposed references to be provided: None Learning Objective: 7 from M-8107L-007 (Main Steam), Rev. 20 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 180 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 61 Points: 1.0 The plant is at rated conditions with the Electrical Pressure Regulator (EPR) in service.

Given the following:

• The EPR Control Position is set at 910 PSI

• The MPR Handwheel Position indicates 920 PSI An overcurrent condition has caused the running EPR oil pump supply breaker to trip open and the Standby EPR oil pump failed to start.

Given the above information, which of the following is correct?

A. Control valves will open, Main Steam pressure will lower; a reactor scram will occur.

B. Control valve closure will result in a reactor scram.

C. Control valves will initially throttle close, and then throttle open; RPV pressure will stabilize slightly higher.

D. Control valves will initially throttle open, and then throttle close; RPV pressure will stabilize slightly lower.

Monticello NRC Written Exam Page: 181 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 61 ANSWER EXPLANATION Answer C DISTRACTOR ANALYSIS General precautions; during power operation, changing the controlling pressure regulator setpoint will have a small but noticeable effect on core thermal power (approximately 3MWt per 10 psig change). Both EPR oil pumps A & B are powered by MCC-142A, so loss will cause the MPR to be in control, reactor pressure will rise and reactor power would be expected to increase ~ 3 MWt, resulting in increased steam flow to the turbine.

A. Incorrect; plausible it examinee believes that the pressure set point fails to 0 psi upon loss of the EPR oil pumps; set point fails high on loss the oil pumps.

B. Incorrect; Plausible if the applicant believes that the EPR remains in control; MPR takes control when EPR output falls below MPR output.

C. Correct D. Incorrect; this is the opposite effect of the correct answer and is plausible if examinee believes that the pressure set point fails to a slightly lower value (e.g. failed to 900 psi).

QUESTION BACKGROUND Tier: 2 Group: 2 K/A: 241000 Reactor/Turbine Pressure Regulator K3.05 Knowledge of the effect that a loss or malfunction of the REACTOR/TURBINE PRESSURE REGULATING SYSTEM will have on following:

Main Turbine Steam Flow Importance Rating: 3.7 10 CFR Part 55: 41.7 10 CFR 55.43.b  : N/A K/A Match: The applicant is asked to identify the plant response to a failure of the EPR pumps.

The plant responds by increasing pressure and steam flow to the main turbine as control changes to the MPR. Turbine steam flow is directly proportional to TCV position.

Monticello NRC Written Exam Page: 182 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 61 SRO Justification: N/A Technical

References:

Ops Man B.05.09-05 Proposed references to be provided: None Learning Objective: M-8107L-048 LO 9 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: MNGP 2010 Audit Q 60 Comments:

Monticello NRC Written Exam Page: 183 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 62 Points: 1.0 A plant startup is continuing following completion of Turbine-Generator roll to 1800 rpm. Prior to Turbine-Generator synchronization, turbine speed lowers and stabilizes at 1550 rpm.

Without further operator action what is an anticipated response, if any?

A. 7-B-33 (TURBINE VIBRATION HIGH) will alarm.

B. Auxiliary Oil Pump will auto start on low oil pressure.

C. No response, this is an expected operating characteristic.

D. 7-B-31/32 (TURB DIFF EXPANSION LONG/SHORT ROTOR) will alarm.

Monticello NRC Written Exam Page: 184 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 62 ANSWER EXPLANATION Answer: A and B DISTRACTOR ANALYSIS A. Incorrect: this is not considered a vibration-sensitive speed; continuous operation near the turbines critical speeds of 1150, 1200 and 1400 rpm is not permitted (acceleration should be constant in these regions), as vibrations are expected to be a concern A. Correct: The flow of steam through the turbine helps to maintain even heating of the internal components. It takes almost no steam flow to maintain the unloaded Turbine-Generator at set speed. Without the benefit of the added steam flow that comes with increasing the load, increased vibration can occur due to rubs and localized heating of Turbine components.

Since the question stem did not mention any time frame associated with operation of the Turbine-Generator at the reduced speed, high vibrations is a likely outcome.

B. Correct: The turbine is designed for 1800 rpm. If turbine speed is less than 1600 rpm the Main Shaft Oil Pump will be ineffective and cannot supply the proper oil requirements so the Auxiliary Oil Pump will start at this time (AOP is placed in AUTO during startup when shaft speed gets above 1600 rpm).

C. Incorrect: decreasing to 1550 rpm is NOT an expected operational occurrence, the normal operation of the turbine is 1800 rpm.

D. Incorrect: a DECREASE in the turbine speed should NOT affect the general precaution temperature limit (HP turbine first stage bowl temperature rate of change should NOT exceed 150°F/hr) so differential expansion is not a concern; differential expansion is more of a concern during startup, QUESTION BACKGROUND Tier: 2 Group: 2 K/A: (245000 Main Turbine Gen. / Aux.) A1.02 Ability to predict and/or monitor changes in parameters associated with operating the MAIN TURBINE GENERATOR AND AUXILIARY SYSTEMS controls including: (CFR: 41.5 / 45.5) Turbine speed Importance Rating: 2.6 10 CFR Part 55: 41.5 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants ability to monitor turbine speed and predict the impacts a decrease in turbine speed will have on the Main Turbine Generator and Auxiliary Systems Monticello NRC Written Exam Page: 185 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 62 SRO Justification: N/A Technical

References:

Ops Man B.06.01-02 and -05, M-8107L-013 (Main Turbine) Rev. 16, M-81 07L-087 (Turbine Auxiliaries) Rev. 10 Proposed references to be provided: None Learning Objective: 9 from M-81 07L-087 (Turbine Auxiliaries) Rev. 10 Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 186 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 63 Points: 1.0 The plant was operating at rated conditions when a station power transfer occurred from the 2R to the 1R transformer. A reactor scram was inserted and all systems responded as expected.

Which one of the following actions must be taken next to control Reactor Feedwater IAW procedure C.4-A (REACTOR SCRAM)?

A. Prepare and restart #11 Reactor Feed Pump.

B. Prepare and restart #12 Reactor Feed Pump.

C. Control RPV water level using HPCI and/or RCIC.

D. Control RPV water level with the Low Flow Reg valve.

Monticello NRC Written Exam Page: 187 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM ANSWER EXPLANATION Question Number: 63 Answer D DISTRACTOR ANALYSIS On a station power transfer, all plant busses will transfer to the 1R transformer with the exception of 12 bus. This is to prevent a second feed pump from starting and complicating the electrical situation. 12 Reactor Feed pump is powered from 12 bus so it will not be running. 11 Reactor Feed pump will remain running and will be sufficient to maintain level post scram. Therefore RPV level will be maintained on the low flow feedwater regulating valve.

A. Incorrect, assuming the power transfer occurs correctly, 11 RFP will remain running.

B. Incorrect, this pump can not be used as 12 Bus will be de-energized. If 11 Bus didnt re-power correctly then 12 Bus would get re-powered and 12 RFP would start.

C. Incorrect, C.4-A action to take if the RFPs arent available.

D. Correct QUESTION BACKGROUND Tier: 2 Group: 2 K/A: 259001 Reactor Feedwater A4.01 Ability to manually operate and/or monitor in the control room: System flow Importance Rating: 3.6 10 CFR Part 55: 41.7 10 CFR 55.43.b  : N/A K/A Match: KA is matched by having the applicant evaluate a plant transient that results the need for the operator to take manual control of system flow.

SRO Justification: N/A Monticello NRC Written Exam Page: 188 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 63 Technical

References:

Ops Man B.09.06-01, C.4-A Proposed references None to be provided:

Learning Objective: MT-ILT-AOP-001L LO 8 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: 2013 MNGP NRC exam Q 61 Comments:

Monticello NRC Written Exam Page: 189 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 64 Points: 1.0 The plant is at rated conditions when the following annunciators alarmed:

• 4-A-12 (OFF GAS HI RADIATION)

• 4-A-2 (OFF-GAS TIMER ACTUATED)

Two minutes later, A & B Off-gas radiation monitors have been verified to be reading at 2 times the alarm setpoint.

Which of the following actions is correct for the conditions above?

A. Immediately Scram IAW C.4-K (IMMEDIATE REACTOR SHUTDOWN) because off-site release has exceeded the limits of C.5-1400 (RADIOACTIVE RELEASE CONTROL).

B. Reduce Reactor Recirculation System flow to minimum then initiate a Manual Reactor SCRAM, in accordance with C.4-K (IMMEDIATE REACTOR SHUTDOWN), because the Off-Gas system has isolated and Main Condenser vacuum is degrading.

C. Reduce power, in accordance with C.4-F (RAPID POWER REDUCTION), to attempt to bring the gross gamma activity rate within limits.

D. Reduce reactor power, in accordance with C.4-F (RAPID POWER REDUCTION), because the Off-Gas system has isolated and Main Condenser vacuum is degrading.

Monticello NRC Written Exam Page: 190 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 64 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A. Incorrect: the signal is VALID and UPSCALE but is well below EOP C.5-1400 entry condition. See C B. Incorrect: the reactor should only be shutdown AFTER the 30-minute timer times out; C.4-K (Immediate Reactor Shutdown) is NOT the first abnormal procedure to enter, C.4-K entry is a secondary entry if the 30-minute timer times out OR activity remains high after the C.4-F (Rapid Power Reduction) procedure; Main Condenser vacuum will not degrade until the timer timers out resulting in recombiner trains trip.

C. Correct: Off-Gas Pretreatment Rad Monitors monitor the short-lived radioisotopes, thus there is 30-minute time delay to validate the condition and before a reactor scram shall occur. Per ARP 4-A-2, if Off-Gas Pretreatment Radiation Monitors indicate a valid UPSCALE, check the MSL rad monitors, notify shift supervisor, and reduce power before 30 minutes elapses using Rapid Power Reduction (C.4-F). When the 30-minute timer times out the recombiner trains trips, which leads to the offgas isolation and eventually to main condenser vacuum degrading.

D. Incorrect: wrong reason; the power should be reduced, but main condenser vacuum does NOT degrade until after the 30-minute timer times out and sometime after offgas recombiner trains trip.

QUESTION BACKGROUND Tier: 2 Group: 2 K/A: (271000 Offgas) 2.4.4 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Importance Rating: 4.5 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A Monticello NRC Written Exam Page: 191 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 64 K/A Match: K/A is matched by testing the applicants ability to recognize abnormal indications and annunicator alarms associated with the Offgas System that are entry conditions for abnormal operating procedure, Ops Man C.4-F SRO Justification: N/A Technical

References:

Ops Man B.05.11-01 and -02, ARP 4-A-2 and 4-A-12, M-8107L-091 (Off-Gas Recombiners) Rev. 11, M-8107L-077 (Process Radiation Monitoring and Area Radiation Monitors) Rev. 15 Proposed references to be provided: None Learning Objective: 2.d and 7 of M-8107L-077 (Process Radiation Monitoring And Area Radiation Monitors) Rev. 15 and 7 of M-8107L-091 (Off-Gas Recombiners) Rev. 11 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 192 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 65 Points: 1.0 The plant is at rated conditions with the following conditions:

• RM-17-452A [Reactor Building exhaust vent plenum monitor] reads 30 mrem/hr

• RM-17-453A [Refueling floor process monitor] reads 33 mrem/hr Which one of the following is correct concerning the A SBGT system?

A. NEITHER RM-17-452A nor RM-17-453A have generated a high signal; therefore A SBGT remains in Standby.

B. ONLY RM-17-452A has generated a high signal; therefore A SBGT starts.

C. ONLY RM-17-453A has generated a high signal; therefore A SBGT starts.

D. BOTH RM-17-452A and RM-17-453A have generated a high signal, EITHER of which starts the A SBGT.

Monticello NRC Written Exam Page: 193 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 65 ANSWER EXPLANATION Answer B DISTRACTOR ANALYSIS SBGT initiation is provided automatically when the following parameters exceed preset limits:

1) Reactor Building ventilation plenum high radiation, one high (26 mrem/hr) or two downscale trips of the two Reactor Building exhaust vent plenum monitors (RM-17-452A and RM 452B), or

2) Refueling floor radiation, one high (50 mrem/hr) or two downscale trips of the two refueling floor process monitors (RM-17-453A and RM-17-453B)

A. Incorrect, start logic is met due to RM-17-452A B. Correct C. Incorrect, setpoint is not met for RM-17-453A D. Incorrect, setpoint is not met for RM-17-453A QUESTION BACKGROUND Tier: 2 Group: 2 K/A: 288000 Plant Ventilation K4.01 Knowledge of PLANT VENTILATION SYSTEMS design feature(s) and/or interlocks which provide for the following: Automatic initiation of standby gas treatment system Importance Rating: 3.7 10 CFR Part 55: 41.7 10 CFR 55.43.b: N/A K/A Match: KA is matched by giving the applicant a set of plant conditions and asking if the A SBGT system status has changed and the reason for the change, if any.

Monticello NRC Written Exam Page: 194 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 65 SRO Justification: N/A Technical

References:

Ops Man B.05.11-02 Proposed references to be provided: None Learning Objective: M-8107L-008 LO 7 Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 195 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 66 Points: 1.0 A licensed reactor operator last completed their 10 CFR 55.53 required medical examination on September 15, 2016.

Which of the following is the latest that they may complete their next medical examination and still meet the requirements of 10 CFR 55.53?

A. September 15, 2017 B. September 15, 2018 C. September 30, 2018 D. December 31, 2018 Monticello NRC Written Exam Page: 196 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 66 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A, B, and D are incorrect; plausible interpretations of 10 CFR 55. Medical physicals are typically recommended yearly in a personal instance, but licensees are required to take physicals every 2 years. In addition, the notification timeline to the Commission for many licensee actions/events varies from minutes to multiple months (e.g., per 10 CFR 50.73, a LER is required to be submitted within 60 days).

C is correct: As stated in 10 CFR 55, Subpart C - Medical Requirements, A licensee shall have a medical examination by a physician every two years and in Subpart F, § 55.53(i) The licensee shall have a biennial medical examination. NUREG 1021 clarifies the biennial requirement to be on the last day of anniversary month in the second year.

QUESTION BACKGROUND Tier: 3 Group: 1 K/A: Generic 2.1.4, Knowledge of individual licensed operator responsibilities related to shift staffing, such as medical requirements, no-solo operation, maintenance of active license status, 10CFR55, etc. (CFR: 41.10 / 43.2)

Importance Rating: 3.3 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the licensees responsibilities related to 10 CFR 55 medical requirements.

SRO Justification: N/A Technical

References:

10 CFR 55.53 (2016); NUREG 1021 (Rev 10); OWI-01.08 (Rev 22)

Proposed references to be provided: None Monticello NRC Written Exam Page: 197 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 66 Learning Objective: N/A Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 198 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 67 Points: 1.0 IAW AWI-04.04.02 (EQUIPMENT POSITIONING, INDEPENDENT AND CONCURRENT VERIFICATION METHODS):

What is the MINIMUM permission required to perform a concurrent verification instead of an independent verification when hanging a Danger Tag on a safety related component?

(Assume significant radiation, safety or other hazards do NOT exist)

A. Control Room Supervisor B. Shift Manager C. Operations Manager D. Plant Manager Monticello NRC Written Exam Page: 199 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 67 ANSWER EXPLANATION Answer C DISTRACTOR ANALYSIS 4 AWI-04.04.02 [equipment positioning, independent and concurrent verification methods] states that independent verification of equipment isolation SHALL be performed for proper component position upon installation of all DANGER cards, The Operations Managers permission is required to perform a concurrent verification for an independent verification when hanging Danger Cards.

C is correct A, B and D are incorrect: Plausible misconceptions on authority QUESTION BACKGROUND Tier: Group:

K/A: Conduct of Operations 2.1.29 Knowledge of how to conduct system lineups, such as valves, breakers, switches, etc.

Importance Rating: 4.1 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: The KA is matched by having the applicant evaluate whose authority is required to deviate from the requirements of independent verification given a specific scenario.

SRO Justification: N/A Technical

References:

4 AWI-04.04.02 Proposed references to be provided: None Monticello NRC Written Exam Page: 200 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 67 Learning Objective:

Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 201 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 68 Points: 1.0 A Breaker and Relay Technician called the Control Room to request that Operations cycle a 4160 VAC breaker for testing using the control switch in the Control Room. Prior to cycling the breaker, the operator finds both a BREAKER/RELAY TEST STATUS card and a Danger tag on the breaker hand-switch.

Which one of the following CORRECTLY identifies the correct operator action(s) for this situation?

A. Obtain verbal assurance from the Breaker and Relay Technician that the breaker is in the test position, then cycle the breaker.

B. Inform the Breaker and Relay Technician that the breaker cannot be cycled until the Danger tag is cleared from the hand-switch.

C. Temporarily remove the Danger tag after receiving verbal authorization from the CRS, cycle the breaker, then replace the Danger tag.

D. Cycle the breaker per the Breaker and Relay Technician's request since the BREAKER/RELAY TEST STATUS card overrides the Danger tag.

Monticello NRC Written Exam Page: 202 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 68 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS FP-OP-TAG-01, section 4.15 - Danger Tags are used to provide worker protection. At no times may a Danger Tagged component be operated except for component position verification by the independent verifier.

In accordance with 4 AWI-04.04.09: Section 4.1.2.B.3 If a Danger or Caution tag and a BREAKER/RELAY TEST STATUS card are placed on the same control switch, the Danger or Caution tag SHALL take precedence and operation of the control switch is prohibited until the Danger or Caution tag is removed.

A. Incorrect; Danger tag must be cleared IAW the tagging procedure before the breaker can be cycled.

B. Correct; Danger tag must be cleared IAW the tagging procedure before the breaker can be cycled.

C. Incorrect; Danger tag must be cleared IAW the tagging procedure.

D. Incorrect; the Danger Tag takes precedence over the test tag.

QUESTION BACKGROUND Tier: 3 Group: 1 K/A: Generic 2.1.26 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen). (CFR:

41.10 / 45.12)

Importance Rating: 3.4 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match:

SRO Justification: N/A Monticello NRC Written Exam Page: 203 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 68 Technical

References:

FP-OP-TAG-01, section 4.15; 4 AWI-04.04.09: Section 4.1.2.B.3 Proposed references to be provided: None Learning Objective: M8108L-039 Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank X Bank _

Question History: Modified from MNGP Exam Bank Question 1551 (M8108L-039 50)

Comments:

Monticello NRC Written Exam Page: 204 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 69 Points: 1.0 The reactor is shutdown for a scheduled refueling outage. The following conditions exist:

• RPV water level is 57 inches

• Reactor Coolant Temperature is 149°F

• The Reactor Mode switch is in SHUTDOWN

• ALL Reactor Vessel Head closure bolts are detensioned

• Division 2 RHR/RHRSW pumps are in Shutdown Cooling (SDC)

• Division 1 AC Busses are deenergized for a scheduled maintenance window Maintenance personnel assigned to work on EDG 11 have instead pumped down the lube oil sump for EDG 12.

Which one of the following Technical Specification LCOs would be NOT MET?

A. LCO 3.4.8 (RHR Shutdown Cooling System - Cold Shutdown).

B. LCO 3.8.2 (AC Sources - Shutdown).

C. LCO 3.8.8 (Distribution Systems - Shutdown)

D. LCO 3.9.8 (RHR - Low Water Level).

Monticello NRC Written Exam Page: 205 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 69 ANSWER EXPLANATION Answer B DISTRACTOR ANALYSIS The maintenance activity has resulted in EDG 12 being inoperable. The unavailability of EDG 12 along with the Division 1 maintenance window, the examinee must determine that no EDGs are operable and LCO 3.8.2 is not satisfied. Then the examinee must also recognize that the plant is in Mode 5.

A. Incorrect, bus 16 remains energized from offsite power and the requirements for LCO 3.4.8 are met.

B. Correct; EDG 12 is no longer available to supply Bus 16.

C. Incorrect; the required distribution system is unaffected and remains energized from offsite power.

D. Incorrect; Bus 16 remains energize from offsite power and the requirements for LCO 3.9.8 are met.

QUESTION BACKGROUND Tier: 3 Group:

K/A: Equipment Control 2.2.36 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Importance Rating: 3.1 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: The KA is matched SRO Justification: N/A Monticello NRC Written Exam Page: 206 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 69 Technical

References:

1.1 Table 1.1-1 TS 3.9.8 Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: MNGP 2013 NRC exam question 9 Comments:

Monticello NRC Written Exam Page: 207 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 70 Points: 1.0 While verifying control switch tags in the Main Control Room, the independent verifier discovered that a control switch was in a different position than that specified on the tag.

Which one of the following action(s) is required to be done FIRST?

A. Initiate an Action Request.

B. Stop and notify the Control Room Supervisor.

C. Verify the required position in the Clearance Order.

D. Reposition the control switch to match the tagged position.

Monticello NRC Written Exam Page: 208 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 70 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A and C are incorrect: These would be plausible actions after the CRS is notified.

B is correct: The individual performing the independent verification or concurrent verification check SHALL NOT physically reposition the component. If the component is not found in the specified condition, the individual SHALL notify the CRS for corrective action.

D is incorrect: This action is specifically not allowed initially.

QUESTION BACKGROUND Tier: 3 Group: 2 K/A: Generic 2.2.13, Knowledge of tagging and clearance procedures. (CFR: 41.10 /

45.13)

Importance Rating: 4.1 10 CFR Part 55: 41.10 10 CFR 55.43.b  : N/A K/A Match: K/A is matched by testing the applicants knowledge of tagging (control switch tags).

SRO Justification: N/A Technical

References:

AWI-04.04-02 section 4.3.6, Proposed references to be provided: None Monticello NRC Written Exam Page: 209 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 70 Learning Objective: 2 from M8108L-039 Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank _

Bank X Question History: 2009 MNGP Audit Written Exam - Question 70 Comments:

Monticello NRC Written Exam Page: 210 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 71 Points: 1.0 The reactor is operating in MODE 2 at approximately 3% Rated Thermal Power (RTP). HPCI is in operation for surveillance testing.

Which of the following is the HIGHEST Torus water temperature that if reached, would still allow the continuance of HPCI surveillance testing IAW Technical Specifications?

A. 89°F B. 94°F C. 99°F D. 109°F Monticello NRC Written Exam Page: 211 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 71 ANSWER EXPLANATION Answer C DISTRACTOR ANALYSIS While monitoring torus temperature the examinee must determine that with torus temperature

>100F a nd pe rform a nce of te s ting tha t a dds he a t to the torus , a ll te s ting tha t a dds he a t to the suppression pool must be immediately suspended.

A is incorrect: This is plausible as it is approaching the EOP entry temperature limit (90°F) and the highest allowed limit if no testing were in progress.

B is incorrect: This is plausible if the candidate believes that the limit is to suspend testing above 95°F.

D is incorrect: This is plausible as it is approaching the scram limit for torus water temperature (110°F).

QUESTION BACKGROUND Tier: 3 Group:

K/A: Equipment Control 2.2.12 Knowledge of surveillance procedures.

Importance Rating: 3.7 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: The KA is matched by having the applicant evaluate TS Surveillance requirements at the RO level Question Number: 71 SRO Justification: N/A Technical

References:

TS 3.6.2.1 Proposed references to be provided: None Learning Objective: M8107L-002 LO 10 Monticello NRC Written Exam Page: 212 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: 2013 MNGP ILT NRC Exam Comments:

Monticello NRC Written Exam Page: 213 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 72 Points: 1.0 The plant is operating at full power with the following conditions:

• Annunciator 4-A-11, REACTOR BUILDING HI RADIATION, is in alarm.

• Area Radiation Monitor, RCIC EQUIPMENT AREA (896), amber HI light is lit.

• Annunciator 259-A-6, RBV EFFLUENT HIGH RADIATION, is in alarm.

• The mid-range detector for Reactor Building Vent WRGM is now providing the input for calculating effluent release.

Which of the following actions is immediately required by procedure for the stated conditions?

A. SCRAM the Reactor, per C.4-A (Reactor Scram).

B. Commence a normal Reactor shutdown, per C.4-K (Immediate Reactor Shutdown).

C. Evacuate all personnel from the Reactor Building.

D. Dispatch the Reactor Building operator to investigate.

Monticello NRC Written Exam Page: 214 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 72 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A & B are incorrect but plausible interpretations of immediate required actions as they are contingent steps later in the AOPs.

C is correct: With the above stated conditions a steam leak exists in the Reactor Building (Secondary Containment). The immediate action of AOP C.4-B.02.04.A (STEAM LEAKS OUTSIDE PRIMARY CONTAINMENT) is to evacuate all unnecessary personnel from the Reactor Building.

D is incorrect: plausible since Annunciator 4-A-11 requires that RP personnel permission is required for entry into the Reactor Building.

QUESTION BACKGROUND Tier: 3 Group: 3 K/A: Generic 2.3.13 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. (CFR: 41.12 / 43.4 / 45.9 / 45.10)

Importance Rating: 3.4 10 CFR Part 55: 41.12 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the response to radiation monitor alarms SRO Justification: N/A Monticello NRC Written Exam Page: 215 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 72 Technical

References:

AOP B.02.04.A, ARP 4-A-11, AWI-08.04.02 Proposed references to be provided: None Learning Objective: MT-ILT-AOP-005L 2 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: Question 1751 of Exam BANK Comments:

Monticello NRC Written Exam Page: 216 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 73 Points: 1.0 The control room has been notified of two unconscious individuals in a locked high-radiation area

[LHRA] / highly contaminated area [HCA] in the Reactor Building. The fire brigade has been assembled and is responding to this serious injury involving the potential loss of life.

Which one of the following completes the statements below?

One member of the fire brigade has lost dosimetry in route to the radiologically controlled area

[RCA], IAW A.5-100 [on-site medical emergencies] the member __(1)__ enter the RCA and respond to victims location within the LHRA / HCA.

IAW A.5-100 the __(2)__ is responsible for contamination control at the accident scene.

A. (1) may (2) radiation protection/chemical technician B. (1) may (2) fire brigade leader C. (1) may not (2) radiation protection/chemical technician D. (1) may not (2) fire brigade leader Monticello NRC Written Exam Page: 217 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 73 ANSWER EXPLANATION Answer A DISTRACTOR ANALYSIS In all cases the health and safety of the injured person(s) is the primary concern. Prompt medical treatment takes precedence over radiological decontamination, radiological monitoring or contamination control methods when serious injury involving the potential loss of life or limb has occurred.

Radiation Protection/Chemical Technicians are responsible for contamination control at the accident scene and radiological monitoring of the victim and emergency response personnel.

A. (1) correct, in this circumstance prompt treatment takes precedence over radiological monitoring (2) correct, the rad pro / chem tech has responsibility for contamination control B. (1) correct, in this circumstance prompt treatment takes precedence over radiological monitoring (2) incorrect, if this was a plant fire the fire brigade leader would be responsible for giving the brigade instructions as to any radiation and/or contamination problems.

C. (1) incorrect, in this circumstance prompt treatment takes precedence over radiological monitoring (2) correct, the rad pro / chem tech has responsibility for contamination control D. (1) incorrect, in this circumstance prompt treatment takes precedence over radiological monitoring (2) incorrect, if this was a plant fire the fire brigade leader would be responsible for giving the brigade instructions as to any radiation and/or contamination problems.

QUESTION BACKGROUND Tier: 3 Group:

K/A: Radiation Control 2.3.12 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.

Importance Rating: 3.2 Monticello NRC Written Exam Page: 218 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 73 10 CFR Part 55: 41.12 10 CFR 55.43.b: N/A K/A Match: The KA is matched having the applicant evaluate a scenario where a plant medical emergency is taking place, during this event normal radiological requirements are suspended if there is a potential for loss of life, the licensed operator has a responsibility as a fire brigade chief to understand the responsibilities of that position as they apply to radiological conditions.

SRO Justification: N/A Technical

References:

A.5-100 [on-site medical emergencies]

Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 219 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 74 Points: 1.0 Which of the following requires the assigned individual(s) to be respirator qualified?

1. Serving as a Fire Brigade member

2. Manning the Control Room during a Toxic Gas event

3. Manning the ASDS Panel for Shutdown Outside the Control Room

4. Manning the Control Room during ALL Emergency Plan implementations A. 1 AND 2 ONLY B. 1, 2, AND 3 ONLY C. 1, 2, AND 4 ONLY D. 1, 2, 3, AND 4 Monticello NRC Written Exam Page: 220 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 74 ANSWER EXPLANATION Answer: A DISTRACTOR ANALYSIS A is correct: Fire Brigade and Toxic gas events are on-shift collateral duties that require respirator qualifications B, C, and D are incorrect: 3 and 4 are not required; however they are both plausible due to some overlapping responsibilities between the four choices.

QUESTION BACKGROUND Tier: 3 Group: 4 K/A: Generic 2.4.12 Knowledge of general operating crew responsibilities during emergency operations. (CFR: 41.10 / 45.12)

Importance Rating: 4.0 10 CFR Part 55: 41.10 10 CFR 55.43.b: N/A K/A Match: K/A is matched by testing the applicants knowledge of the operating crew responsibility during emergency operations, particularly related to an required emergency devices in different circumstances/locations SRO Justification: N/A Technical

References:

Operations Work Instruction OWI-01-06 (Rev 53)

Proposed references to be provided: NONE Monticello NRC Written Exam Page: 221 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 74 Learning Objective:

Cognitive Level: Higher _

Lower X Question Source: New _

Modified Bank _

Bank X Question History: Question 74 from 2009 MNGP ILT Exam Comments: The first two randomly selected K/As for this question were rejected.

Monticello NRC Written Exam Page: 222 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question 75 Points: 1.0 The plant was at rated conditions when an inadvertent initiation of HPCI occurred. The HPCI Aux Oil Pump was placed in PULL TO LOCK (PTL) IAW C.4-G (INADVERTANT ECCS INITIATION).

Subsequently, a loss of Feed Pumps occurred causing a scram and the following conditions:

• RCIC received an auto initiation and is injecting

• CRD and SBLC are unavailable for injection

• C.5-1100 (RPV CONTROL) has been entered If RPV water level continues to lower, HPCI would be A. AVAILABLE and used for injection.

B. ONLY available for RPV pressure control.

C. LEFT in the PTL condition IAW C.4-G.

D. ONLY used if Torus water level is maintained above -3.3 feet.

Monticello NRC Written Exam Page: 223 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 75 ANSWER EXPLANATION Answer A DISTRACTOR ANALYSIS HPCI should have auto initiated when RCIC did, and with RPV level continuing to lower, HPCI should be used to attempt to restore level. The requirements of the EOPs are to be followed if there is any conflicting guidance between them and the C.4s. In this case the AOP procedure allows the use of HPCI as a mitigating strategy for EOP level control.

B is incorrect: HPCI would be available for injection or pressure control.

C is incorrect: The requirements of the EOPs are to be followed if there is any conflicting guidance between them and the C.4s.

D is incorrect: HPCI is available for use as long as torus water level is maintained above -3.7 feet or needed for core cooling.

QUESTION BACKGROUND Tier: 3 Group:

K/A: Emergency Procedures / Plan 2.4.8 Knowledge of how abnormal operating procedures are used in conjunction with EOPs Importance Rating: 3.8 10 CFR Part 55: 41.12 10 CFR 55.43.b: N/A K/A Match: The KA is matched by knowing the AOP procedure allows the use of HPCI as a mitigating strategy for EOP level control.

SRO Justification: N/A Monticello NRC Written Exam Page: 224 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC RO WRITTEN EXAM Question Number: 75 Technical

References:

Ops Man C.4-G, C.5-1-1100, OWI-01.04 Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: 2009 MNGP Audit Exam Comments:

Monticello NRC Written Exam Page: 225 of 225 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 76 Points: 1.0 The plant was at rated conditions when a transient occurred that required a manual reactor scram. All control rods are fully inserted and plant conditions have been stabilized.

The OATC announced receipt of annunciator: 5-A-4 (SRM DOWNSCALE).

The OATC reported SRM indications as follows:

Channel 21 Channel 22 Channel 23 Channel 24 8 cps 7 cps 4 cps 6 cps Given the above reports and indications, which one of the following describes the CRS recommended action(s)?

A. BYPASS SRM 23 then initiate a work order for SRM Channel 23.

B. Enter and initiate the actions of C.4-B.09.13.B (LOSS OF Y-20).

C. Enter and initiate the actions of C.4-B.09.11.A (LOSS OF A 24 VDC BUS).

D. Insert the SRM and IRM detectors per C.4-A Part A (REACTOR SCRAM).

Monticello NRC Written Exam Page: 1 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 76 Answer Explanation Answer: D DISTRACTOR ANALYSIS A. Incorrect: Not required unless detectors are already driven in B. Incorrect: Drive power is on therefore Y-20 Ckt 5 is available C. Incorrect: These are not indications of a loss of detector power, lights are illuminated on the panel D. Correct: Step 9 of PART A of C.4-A directs the OATC to select drive in on the SRM-IRM detector drive control (on the SRM/IRM panel) to allow the detectors to be driven in. The detectors need to be in the core to more accurately monitor reactor power.

Question Background Tier: 1 Group: 1 K/A: (295006 SCRAM / 1) AA2.01 Ability to determine and/or interpret the following as they apply to SCRAM: (CFR: 41.10 / 43.5 / 45.13) Reactor power Importance Rating: 4.6 10 CFR Part 55: 41.10 / 43.5 / 45.13 10 CFR 55.43.b: 5 K/A Match: K/A is matched by testing the applicants ability to determine reactor power after a SCRAM has occurred and interpret what steps should be taken to get a better understanding of reactor power SRO Justification: involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed Technical

References:

C.4-A, ARP 5-A-4, and M-8114L-001 Proposed references to be provided: Photo that was included in Question 77 of the 2009 Audit Exam (included above)

Monticello NRC Written Exam Page: 2 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 76 Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: Question 77 of 2009 Audit Exam Comments: Question modified slightly, but essentially the same question.

Monticello NRC Written Exam Page: 3 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 77 Points: 1.0 The plant was at rated conditions when an event occurred that required IMMEDIATE Control Room evacuation IAW C.4-C (SHUTDOWN OUTSIDE CONTROL ROOM).

Fifteen (15) minutes after abandoning the Control Room, operators are unable to place the Master ASDS Transfer Switch in the TRANSFER position.

For the above conditions; which of the following describes (1) the control board manipulations required to be taken prior to leaving the control room?

(2) the emergency classification for the event?

A. (1) Depress the manual scram pushbuttons and leave the Rx Mode Switch in RUN (2) declare an ALERT B. (1) Depress the manual scram pushbuttons and leave the Rx Mode Switch in RUN (2) declare a SITE AREA EMERGENCY C. (1) Depress the manual scram pushbuttons and place the Rx Mode Switch in SHUTDOWN, (2) declare an ALERT D. (1) Depress the manual scram pushbuttons and place the Rx Mode Switch in SHUTDOWN, (2) declare a SITE AREA EMERGENCY Monticello NRC Written Exam Page: 4 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 77 Answer Explanation Answer B DISTRACTOR ANALYSIS C-4-C states: IF immediate Control Room evacuation is required, THEN: Manually Scram the Reactor, Leave the mode switch in RUN, Obtain the ASDS keyset from the Shift Supervisors Office key cabinet, Proceed to the ASDS Panel on the 3rd floor of EFT Building.

Leaving the Mode Switch in Shutdown is plausible as this is done for all other scrams. However, this procedure requires it to be left in RUN.

Entry into C.4-C, Shutdown Outside the Control Room, for Control Room evacuation would require declaration of and ALERT but given that 15 minutes have elapsed and control has not been established at the alternate shutdown panel this inability to establish plant control from outside the Control Room will escalate this event to a SITE AREA EMERGENCY.

A. Incorrect, plausible but wrong EAL selection B. Correct C. Incorrect, plausible but wrong mode switch position and EAL selection D. Incorrect, plausible but wrong mode switch position Question Background Tier: 1 Group: 1 K/A: 295016 Control Room Abandonment 2.4.30 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.

Importance Rating: 4.1 10 CFR Part 55: 41.10 10 CFR 55.43.b: 1, 5 Monticello NRC Written Exam Page: 5 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 77 K/A Match: The KA is matched by having the applicant evaluate the actions to take prior to exiting the control room and classification of the event.

SRO Justification: Abnormal action determination and EAL declaration Technical

References:

Ops Man C.4-C A.2-101, Classification of Emergencies Proposed references to be provided: Form 5790-101-02, MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY ACTION LEVEL MATRIX Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: 2010 MNGP ILT NRC Written Exam Q6, significantly modified to SRO level Comments:

Monticello NRC Written Exam Page: 6 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 78 Points: 1.0 A plant cooldown is in progress using B RHR with #12 Reactor Recirculation pump in service. Given the following:

• At 1500: RPV pressure is 15 psig and steady

• At 1501: 16 Bus receives a lockout due to a bus fault

• At 1601: RPV pressure is 125 psig and rising slowly Which of the following represents the MINIMUM, Technical Specification REQUIRED ACTIONS, which must be performed for this event (Assume: loss of safety function does NOT exist)?

A. (1) LCO 3.4.7 REQUIRED ACTIONS A.1, A.2, AND A.3 (2) LCO 3.8.4 REQUIRED ACTIONS A.1, A.2, AND A.3 (3) LCO 3.8.7 REQUIRED ACTION A.1 B. (1) LCO 3.4.8 REQUIRED ACTION A.1 (2) LCO 3.4.9 REQUIRED ACTIONS A.1 AND A.2 (3) LCO 3.8.8 REQUIRED ACTION A.1 AND A.2 C. (1) LCO 3.4.9 REQUIRED ACTIONS A.1 AND A.2 (2) LCO 3.8.4 REQUIRED ACTIONS A.1, A.2, AND A.3 (3) LCO 3.8.7 REQUIRED ACTION A.1 D. (1) LCO 3.8.1 REQUIRED ACTIONS A.1, A.2, AND A.3 (2) LCO 3.8.4 REQUIRED ACTIONS A.1, A.2, AND A.3 (3) LCO 3.8.7 REQUIRED ACTION A.1 Monticello NRC Written Exam Page: 7 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 78 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS A. Incorrect: IAW LCO 3.0.6, LCO 3.4.7 Condition A is entered but the required actions are NOT require to be performed. Additionally, the allowable plat heat up rate is being exceeded (103F/hr) which requires entry into LCO 3.4.9 A also.

B. Incorrect: Plausibly correct if Mode 4 is used.

C. Correct: Allowed plant heat up is being exceeded (103F/hr) which requires entry into LCO 3.4.9 A. LCO 3.8.7 Condition A is entered due to the lockout of Bus 16, and Condition A also directs entry in LCO 3.8.4 Condition A D. LCO 3.8.1 is still being met, therefore no actions are require to be taken for this LCO.

Plausible if examinee believes that offsite sources are also inoperable due to the bus lockout. Additionally, the allowable plat heat up rate is being exceeded (103F/hr) which requires entry into LCO 3.4.9 A also.

QUESTION BACKGROUND Tier: 1 Group: 1 K/A: (295021 Loss of Shutdown Cooling / 4) 2.1.20 Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)

Importance Rating: 4.6 10 CFR Part 55: CFR: 41.10 / 43.5 / 45.12 10 CFR 55.43.b: 2 K/A Match: K/A is matched by testing the applicants ability to interpret and execute the technical specifications related to the present conditions, specifically the loss of SDC SRO Justification: Facility operating limitations in the TS and their bases Technical

References:

TS 3.0.4, 3.0.6, 3.4.9, 3.8.7, Steam Tables, and MT-OPS-ITS-002L Proposed references to be provided: TS 3.0, 3.4.7, 3.4.8, 3.4.9, 3.8.4, 3.8.1, 3.8.7, 3.8.8, and Steam Tables Monticello NRC Written Exam Page: 8 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 78 Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: Question 79 of 2009 MNGP Audit exam Comments:

Monticello NRC Written Exam Page: 9 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 79 Points: 1.0 A Loss of Coolant Accident has occurred and the following conditions are present:

• Drywell pressure is 32 psig and rising

• Drywell Temperature is 282°F and rising

• RPV pressure is 800 psig and lowering

• RPV water level is -50 inches and rising

• Torus water level is +2 feet and rising

• 16 Bus is LOCKED OUT

• RHR is lined up for LPCI injection

• The Drywell Cooling Fan switches are in OFF

• The Recirc Pumps are tripped As the CRS, which of the following is the next action to be directed based on the above information?

A. Lower Torus level B. Initiate Drywell Sprays C. Initiate a Blowdown due to exceeding 281°F in the Drywell D. Initiate a Blowdown due to exceeding Pressure Suppression Pressure Monticello NRC Written Exam Page: 10 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 79 Answer Explanation Answer B DISTRACTOR ANALYSIS Initial operator actions must be taken to place drywell sprays in service, Torus sprays may already be in service, the decision as to whether or not conditions can stay within the PSP or D/W temp can be restored should not be made until drywell sprays are placed in service and an evaluation can be made as to the expected response.

A. Incorrect, as the flow chart steps must be prioritized and torus level cannot be lowered quickly because MO-2032 (MCC-142) does not have power.

B. Correct C. Incorrect, as the ability to restore and maintain cannot be evaluated prior to initiating sprays. Additionally Blowdown is not required until it is determined that DW Temperate cannot be restored and maintained below 338°F.

D. Incorrect, as the ability to maintain below the limit cannot be evaluated prior to initiating sprays.

Question Pedigree Tier: 1 Group: 1 K/A: 295024 High Drywell Pressure EA2.01 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE: Drywell pressure.

Importance Rating: 4.2 10 CFR Part 55: 41.10 10 CFR 55.43.b: 5 Monticello NRC Written Exam Page: 11 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 79 K/A Match: The KA is matched by having the applicant evaluate the plant conditions and select a proper action during accident conditions.

SRO Justification: The SRO must have detailed knowledge of actions required by the EOP flowchart based on given conditions.

Technical

References:

C.5-1100 and C.5-1200 Proposed references to be provided: C.5-1100 and 1200 with entry conditions blanked out.

Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: MNGP bank 1972 Comments:

Monticello NRC Written Exam Page: 12 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 80 Points: 1.0 A Loss of Normal Off-Site power occurred resulting in the following plant conditions:

• Drywell pressure is 1.9 psig and slowly rising

• Drywell temperature is 150F and slowly rising Assuming a LOCA has NOT occurred; which of the following actions must be directed next to properly mitigate the Primary Containment conditions above?

A. Initiate Torus Sprays.

B. Initiate Drywell Sprays.

C. Lower RPV Pressure 500-700 psig D. Start ALL available Drywell Cooling Monticello NRC Written Exam Page: 13 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 80 Answer Explanation Answer: D DISTRACTOR ANALYSIS A loss of off-site power will result in a reactor trip (manual or automatic depending on timing), loss of both Recirc Pumps, Loss of Circ Water/Vacuum and Main Turbine trip, loss of normal feed, and loss of Drywell Cooling. Examinee must determine that the degrading Drywell parameters are due solely to a loss of Drywell cooling. The ECCS signal should be bypasses and all fans started.

A. Incorrect; required by Containment Control EOP before DW pressure reaches 12 psig, but Drywell pressure increase is simply due to loss of cooling and therefore spraying the Torus would have a negligible effect.

B. Incorrect; required only if unable to maintain DW pressure below 12 psig by other mitigating methods (e.g., restoring DW cooling).

C. Incorrect; lowering RPV pressure to 500-700 psig is a mitigating strategy for a Non-ATWS LOCA.

D. Correct; Restoring DW cooling would have an immediate effect.

Question Background Tier: 1 Group: 1 K/A: (295028 High Drywell Temperature / 5) EA2.01 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL TEMPERATURE: (CFR: 41.10 /

43.5 / 45.13) Drywell temperature Importance Rating: 4.1*

10 CFR Part 55: CFR: 41.10 / 43.5 / 45.13 10 CFR 55.43.b: 5 K/A Match: K/A is matched by testing the applicants ability to interpret the drywell temperature as a high drywell temperature and EOP-entry condition; the applicant must demonstrate ability to distinguish this from other potential EOP parameters.

Monticello NRC Written Exam Page: 14 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 80 SRO Justification: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Technical

References:

Proposed references to be provided: C.5-1200 Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 15 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 81 Points: 1.0 Given the following plant conditions:

• The unit is operating at rated power

• 4-B-4 (SUPPRESSION WATER LEVEL HI/LOW) is received

• Torus water level is -5 inches and lowering Complete the statements below?

The DBA LOCA initial conditions assumed for Torus water level in the Safety Analyses __(1)__.

If water level were to lower until the Downcomer lines were uncovered __(2)__ would be lost.

A. (1) are met (2) pressure suppression function B. (1) are met (2) Torus cooling capability C. (1) are NOT met (2) pressure suppression function D. (1) are NOT met (2) Torus cooling capability Monticello NRC Written Exam Page: 16 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 81 Answer Explanation Answer C DISTRACTOR ANALYSIS The suppression pool volume ranges between 68,000 ft3 at the low water level limit of - 4.0 inches and 72,910 ft3 at the high water level limit of + 3.0 inches. Suppression pool water level must be maintained within the limits specified so that the safety analysis remains valid.

With suppression pool water level outside the limits, the conditions assumed for the safety analyses are not met. If water level is below the minimum level, the pressure suppression function still exists as long as downcomer lines are covered; HPCI and RCIC turbine exhausts are covered, and S/RV quenchers are covered. If suppression pool water level is above the maximum level, protection against overpressurization still exists due to the margin in the peak containment pressure analysis and the capability of the Drywell Spray System.

A. Incorrect, plausible but wrong if the applicant evaluates the alarm condition and level and concludes that the level is within the T.S. limits.

B. Incorrect, plausible but wrong if the applicant evaluates the alarm condition and level and concludes that the level is within the T.S. limits and does not recognize that at the normal flow rates associated with one loop of pool cooling that vortex limits are not exceeded until Torus level drops to -4.5 feet (approximately a foot below the bottom of the downcomer lines.

C. Correct D. Incorrect, plausible but wrong if the applicant does not recognize that at the normal flow rates associated with one loop of pool cooling that vortex limits are not exceeded until Torus level drops to -4.5 feet (approximately a foot below the bottom of the downcomer lines.

Question Background Tier: 1 Group: 1 K/A: 295030 Low Suppression Pool Wtr Lvl 2.4.31 Knowledge of annunciator alarms, indications, or response procedures.

Importance Rating: 4.1 10 CFR Part 55: 41.10 10 CFR 55.43.b: 2 Monticello NRC Written Exam Page: 17 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 81 K/A Match: The KA is matched by having the applicant evaluate an alarm condition and plant indications and apply this information to determine the potential impact on safety analyses SRO Justification: This is an SRO level question because the candidate must have knowledge of TS bases and is beyond above the line knowledge Technical

References:

C.6-004-B-04 Technical Specifications and Bases Proposed references to be provided: None Learning Objective: M-8107L-044, LO 10b Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 18 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 82 Points: 1.0 The plant is in a refueling outage with the following conditions present:

• The RPV head is removed

• RPV water level is two-feet (2) above the reactor vessel flange

• 1027 elevation radiation levels are 2 mrem/hr

• Secondary containment is NOT established A leak occurs in the East Shutdown Cooling room thats confirmed by the Reactor Building Operator and plant conditions changed as follows:

• RPV water level lowered BY 224 inches

• 1027 elevation radiation levels rises to 18 mrem/hr

• 12 minutes after the start of the leak, it was isolated and RPV water level was restored to the flange Given the above information, Figure 28 from Operation Manual B.01.01-06 [following page], and the provided Monticello Nuclear Generating Plant Emergency Action Level Matrix; Which of the following E-plan classifications, if any, should be declared?

A. Notification of Unusual Event B. Alert C. Site Area Emergency D. No Classification required Monticello NRC Written Exam Page: 19 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 82 Monticello NRC Written Exam Page: 20 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 82 Answer Explanation Answer: A DISTRACTOR ANALYSIS A. Correct: The plant is in Mode 5 and a VALID water level lowering is observed. Reactor water level lowers by 224 resulting in indicated water level reading of approximately -25 (applicant must view Figure 28 below, note an original level of approximately 655+2=679, then subtract 224 from that to arrive at 455, which is approximately 25 above the Low-Low Level trip (-47) = -22. An UNPLANNED VALID Area Radiation Monitor rises as indicated by the 1027 elevation rad monitors (however, the rise is not enough to trip an alarm or in an area where continuous occupancy is required). These observations would meet the requirement of RU2.1 in the EAL Matrix.

B. Incorrect: radiation levels on the 1027 elevation would have to reach the alarm setpoint of 20 mrem to meet the requirements of RA2.1, in this case the rad monitors stay <20 mrem at 18mrem.

C. Incorrect: Plausible as secondary containment is NOT established however water level has not gone below - 53 in order to meet CS2.1.

D. Incorrect: Plausible if Hot EALs are used or if CU2.1 is used and less than 15 minutes is not recognized.

Question Background Tier: 1 Group: 1 K/A: (295031 Reactor Low Water Level / 2) EA2.01 Ability to determine and/or interpret the following as they apply to REACTOR LOW WATER LEVEL: (CFR: 41.10 / 43.5 /

45.13) Reactor water level Importance Rating: 4.6*

10 CFR Part 55: CFR: 41.10 / 43.5 / 45.13 10 CFR 55.43.b: 5 K/A Match: K/A is matched by testing the applicants ability to interpret the water level decrease and its application to a low water level reading, then use that information to apply it to a EAL classification SRO Justification: E-Plan EAL Determination Monticello NRC Written Exam Page: 21 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 82 Technical

References:

Ops Man B.01.01-06 Figure 28, E-Plan EAL Matrix, and M-8107L-028 (Reactor Vessel And Assembly) Rev. 15 Proposed references to be provided: Ops Man B.01.01-06 Figure 28 (inserted within question; and E-Plan EAL Matrix Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: Question 80 2010 MNGP ILT and Question 79 of 2009 MNGP ILT - with minor editorial changes and a change to the duration the water level is below the flange (from 14 mins to 12 mins) so there is more margin from the EAL threshold Comments:

Monticello NRC Written Exam Page: 22 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 83 Points: 1.0 A transient occurred resulting in the following conditions:

• Drywell pressure is 58 psig and slowly rising

• Torus water level is 10 ft. and steady

• Drywell temperature is 320°F and slowly rising

• Drywell radiation is 10 R/hr and slowly rising

• Attempts to spray the drywell have been unsuccessful

• Significant fuel damage is anticipated You have determined that it is necessary to vent the Primary Containment in accordance with C.5-3505 (VENTING PRIMARY CONTAINMENT).

Given the above information, which one of the following choices identifies:

(1) The recommended vent path?

(2) The desired strategy for venting Primary Containment?

A. (1) SBGT through the 18 inch torus vent (C.5-3505 PART C)

(2) Venting MUST be limited to ONLY the volume required to maintain pressure below the DW pressure limit.

B. (1) SBGT through the 18 inch torus vent (C.5-3505 PART C)

(2) Venting MAY be extended for a period of time to reduce the amount of radioactivity that may have to be released once fuel damage occurs.

C. (1) Hard Pipe Vent (C.5-3505 PART A)

(2) Venting MUST be limited to ONLY the volume required to maintain pressure below the DW pressure limit.

D. (1) Hard Pipe Vent (C.5-3505 PART A)

(2) Venting MAY be extended for a period of time to reduce the amount of radioactivity that may have to be released once fuel damage occurs.

Monticello NRC Written Exam Page: 23 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 83 Answer Explanation Answer D DISTRACTOR ANALYSIS The preferred method for venting Primary Containment is through the Torus and SBGT so that the discharge will be filtered, scrubbed and elevated. However, if SBGT ductwork is in jeopardy of rupturing due to high pressure in containment, >2.9 psig, then the Hard Pipe Vent should be used to minimize potential impacts on Reactor Building equipment from postulated ductwork failure.

Early or extended Primary Containment pressure reduction to limit radioactivity release may be appropriate if: Significant fuel damage is anticipated. Reducing primary containment pressure while the primary containment atmosphere is still relatively clean increases the capacity of the containment to retain fission products. Later releases, after core damage has progressed, may thereby be avoided.

A. (1) Incorrect, due to potential to impacts on reactor building equipment (2) Incorrect, plausible as in general pressure should only be reduced to maintain below the DW pressure limit, but with significant fuel damage anticipated early or extended releases increase the capacity of containment to retain fission products.

B. (1) Incorrect, see A (1)

(2) Correct C. (1) Correct (2) Incorrect, see A (2)

D. (1) Correct (2) Correct Question Background Tier: 1 Group: 2 K/A: 295010 High Drywell Pressure AA2.03 Ability to determine and/or interpret the following as they apply to HIGH DRYWELL PRESSURE : Drywell radiation levels...

Importance Rating: 3.6 Monticello NRC Written Exam Page: 24 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 83 10 CFR Part 55: 41.10 10 CFR 55.43.b: 5 K/A Match: This question satisfies the K/A statement by asking the applicant to determine which procedure part to use to vent containment based on anticipated fuel damage and that extended venting is warranted SRO Justification: The SRO must have detailed knowledge of procedure parts to determine which vent path is appropriate for the given conditions, this level of procedure knowledge is above the RO knowledge level.

Technical

References:

C.5-3505 C.5-1200 C.5.1-1200 Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 25 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 84 Points: 1.0 A plant startup from cold shutdown is in progress with the following conditions:

• Reactor power is 5 x 103 cps on the SRMs

• The RWM has been bypassed due to a fault

• A second RO is at C-05 to verify control rod movements An administrative error has resulted in several Control Rods being withdrawn that are inconsistent with the Banked Position Withdrawal Sequence (BPWS).

If a Control Rod Drop Accident (CRDA) were to occur under these conditions; complete the following two statements?

Receipt of alarm (1) will validate a Scram resulted from the CRDA.

The Scram (2) ensure that significant fuel damage is prevented.

A. (1) IRM HI HI/INOP (2) WILL B. (1) IRM HI HI/INOP (2) WILL NOT C. (1) APRM UPSC/INOP TRIP (2) WILL D. (1) APRM UPSC/INOP TRIP (2) WILL NOT Monticello NRC Written Exam Page: 26 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 84 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS The power transient will be terminated when sufficient negative reactivity is added to compensate for the positive reactivity added by the dropped rod. The Doppler reactivity coefficient will initially act to limit the power rise, but control rod insertion is necessary to shut down the reactor.

A. Incorrect; the IRM HI HI/INOP scram will terminate the power excursion, but because the BPWS sequence rules were not followed there is no assurance the significant fuel damage will not occur.

B. Correct; IRM HI HI/INOP will terminate the power excursion. Because the BPWS sequence rules were not followed there is no assurance the significant fuel damage will not occur.

C. Incorrect; since the initial power level is in the Source Range the IRM HI HI/INOP scram will terminate the power excursion before the APRM UPSC trip. If power starts out in the upper ranges of the IRM where there may be significant overlap with the APRMs, the APRM UPSC trip may terminate the power excursion before the IRM HI HI trip. Additionally, because the BPWS sequence rules were not followed there is no assurance the significant fuel damage will not occur.

D. Incorrect; see C QUESTION BACKGROUND Tier: 1 Group: 2 K/A: (295014 Inadvertent Reactivity Addition / 1) 2.4.46 Ability to verify that the alarms are consistent with the plant conditions. (CFR: 41.10 / 43.5 / 45.3 / 45.12)

Importance Rating: 4.2 10 CFR Part 55: 41.10 / 43.5 / 45.3 / 45.12 10 CFR 55.43.b: 2 Monticello NRC Written Exam Page: 27 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 84 K/A Match: K/A is matched by testing the applicants ability to verify that the inadvertent reactivity addition (rod drop accident) leads to consistent alarms. The applicant must understand the current situation and make the determination of what alarms would be expected if a rod drop accident were to occur, given the RWM is INOP thus a larger reliance on the operators observations.

SRO Justification: Knowledge of USAR CRDA and/or TS bases are required to select the correct answer.

Technical

References:

Monticello USAR Section 14.7.1; TS BASES; Ops Man B.05.02-05.G.2, B.05.01.02-03, C.1, ARP 5-A-13; M-8107L-001 (Rod Worth Minimizer (RWM)) Rev. 12 Proposed references to be provided: None Learning Objective: 7.a of M-8107L-001 (Rod Worth Minimizer (RWM)) Rev. 12 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History: Used Question 83 from MNGP 2009 ILT as starting point - significantly changed; distractors were rewritten to include annunciator alarms Comments: Question 83 of MNGP 2009 ILT pertained to thermal limits and was an SRO-only question, so with the addition of annunciator alarms this question seems to satisfy SRO-only requirements.

Monticello NRC Written Exam Page: 28 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 85 Points: 1.0 The plant was at rated conditions when a seismic event occurred.

The following conditions exist:

• RPV water level is -100 and slowly lowering

• RPV pressure is 600 psig and slowly lowering

• HPCI is injecting at rated flow

• Both trains of Drywell Sprays are in operation.

• Torus level is +3 feet and stable

• CST level is +3 feet and stable Given the above conditions, which of the following Torus Level mitigating actions is required to be implemented IAW C.5-1200 (PRIMARY CONTAINMENT CONTROL)?

A. Stop RPV injection from the HPCI system B. Drain Torus Water to Radwaste C. Perform Emergency RPV Depressurization D. Stop Drywell Sprays and anticipate RPV Depressurization Monticello NRC Written Exam Page: 29 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 85 Answer Explanation Answer B DISTRACTOR ANALYSIS When the HPCI System is initiated, the pump suction is automatically taken from the condensate storage tank and can be manually or automatically transferred to the suppression pool. The system is initially aligned to take suction from the condensate storage tanks; a preferred source because of improved water quality. The lineup permits faster Reactor restarts following non-accident injections into the reactor vessel. The pump suction will automatically transfer to the suppression pool when the level in the condensate storage tank (as measured from tank bottom) drops below 2-8 in either tank, or if the water in the suppression pool rises 2 above the normal level.

Restore and maintain torus level below +3.0 inches Efforts to restore torus water level to the normal band should be continued while the need for further action is evaluated. Procedure C.5-3402 (DRAINING TORUS WATER TO RADWASTE) provides instructions for reducing torus water level by transferring water to Radwaste through RHR. If torus water level cannot be restored and maintained below 3.7 feet, a blowdown will ultimately be required.

A. Incorrect; Torus level has exceeded the swap over level to align the HPCI suction to the torus, therefore HPCI is no longer taking suction from a source outside the Containment.

B. Correct; Torus water level is above 3 but is relatively stable at 3 and being maintained below 3.7.

C. Incorrect; Torus water level is currently being maintained below 3.7, therefore Blowdown in not required.

D. Incorrect; Torus water level currently being maintained at approximately 3, therefore level is well below the threshold (4.2) for the specified action.

Question Background Tier: 1 Group: 2 K/A: 295029 High Suppression Pool Wtr Lvl EA2.01 Ability to determine and/or interpret the following as they apply to HIGH SUPPRESSION POOL WATER LEVEL : Suppression pool water level...

Importance Rating: 3.9 10 CFR Part 55: 41.10 10 CFR 55.43.b: 5 Monticello NRC Written Exam Page: 30 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 85 K/A Match: The applicant is given plant conditions where based on high torus water level auto swap over of HPCI has occurred and operators are required to use EOPs to mitigate the high level event.

SRO Justification: The SRO must have detailed procedure knowledge to select the appropriate procedure to address high suppression pool water level.

Technical

References:

C.5.1-1200 B.03.02-01 C.5-3402 Proposed references to be provided: C.5-1200 Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 31 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 86 Points: 1.0 The plant has been operating at rated conditions for several months.

On July 6th at 0900 it is discovered that SR 3.5.1.1, with a required frequency of 31 days, has NOT been completed since June 2th at 0800 for the A RHR/LPCI injection subsystem.

Given only the above statement, when is the LATEST the surveillance can be completed for the system status to remain in compliance with Technical Specifications? (Technical Specification Section SR 3.0 is provided.)

A. July 7th at 0900 B. July 11th at 0200 C. August 6th at 0900 D. August 14th at 0200 Monticello NRC Written Exam Page: 32 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 86 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS A and C are incorrect: plausible due to incorrect application of SR 3.0.3: A is 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after discovery and C is 31 days after discovery (these options would be permitted assuming the applicant believed the surveillance was not completed within the specified frequency), but as of July 6th at 0900 it is still within the specified frequency time (with application of the SR 3.0.2 grace period).

D is incorrect: plausible for incorrect combination of SR 3.0.2 and 3.0.3 B is correct: As the conditions state, it has been 34 days since the last successful completion of SR 3.5.1.1. SR 3.5.1.1 has 31 day frequency; however, with the SR 3.0.2 grace period, the required completion date is extended 7.75 days. As such, the surveillance must be completed by 0200 on July 11th.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (203000 RHR/LPCI: Injection Mode) 2.2.12 Knowledge of surveillance procedures.

(CFR: 41.10 / 45.13)

Importance Rating: 4.1 10 CFR Part 55: CFR: 41.10 / 45.13 10 CFR 55.43.b: 2 K/A Match: K/A is matched by testing the applicants knowledge of the surveillance procedures, specifically the SR 3.0 section and its application to TS 3.5.1 (RHR/LPCI injection)

SRO Justification: Facility operating limitations in the TS and their bases; Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7; SR 3.0.1 thru 3.0.4)

Technical

References:

SR 3.0.2, SR 3.0.3, TS 3.5.1, and M-8107L-023 (RHR System) Rev. 20 Proposed references to be provided: SR 3.0.1 thru SR 3.0.4 Monticello NRC Written Exam Page: 33 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 86 Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank X Bank _

Question History: Used Question 96 from 2013 MNGP ILT and Question 2073 from Exam BANK as the starting point but question is significantly modified from those and likely considered NEW Comments: Distractor D is relatively weak; another date could be listed and considered plausible due to applicants miscalculation and/or incorrect use of TS 3.5.1 Monticello NRC Written Exam Page: 34 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 87 Points: 1.0 Given the following plant conditions:

• The plant is at rated conditions

• At 1215 on November 1st, the 11 Core Spray pump tripped while being run for quarterly testing, the control switch was placed in PULL-TO-LOCK

• At 2300 on November 3rd, the 12 Core Spray pump becomes INOPERABLE

• At 0115 on November 4th, the 11 Core Spray pump was declared OPERABLE Including any extensions that are permitted by Technical Specifications, which of the following describes the LATEST time and date the unit must be placed in Mode 3 if the 12 Core Spray pump remains INOPERABLE? [Reference provided T.S. 3.5.1, LCO 1.3 & 3.0]

A. 1215 on November 8th B. 0015 on November 10th C. 2300 on November 10th D. 2300 on November 11th Monticello NRC Written Exam Page: 35 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 87 ANSWER EXPLANATION Answer B DISTRACTOR ANALYSIS A. Incorrect, Refer to correct answer explanation. This answer is plausible if the pumps were associated with the same train in which case the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension time would not be applicable.

B. Correct. In accordance with Section 1 of TS (use and application), when a subsequent train, subsystem, or component expressed in the condition is discovered inoperable or not within limits, the completion time may be extended provided two criteria are met: the subsequent inoperability must exist concurrent with the first inoperability and must remain inoperable or not within limits after the first inoperability is resolved. In this case the more limiting time must be used.

C. Incorrect, This time corresponds with the second inoperability which is the less restrictive time and therefore cannot be used D. Incorrect, This time corresponds with the second inoperability plus the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> extension, which is another improper application.

QUESTION BACKGROUND Tier: 2 Group: 1 K/A: 209001 LPCS A2.01 Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Pump trips Importance Rating: 3.4 10 CFR Part 55: 41.5 10 CFR 55.43.b: 2 K/A Match: This question satisfies the K/A statement by asking the applicant to use T.S. and determine the required time to be in mode 3 based on equipment availability.

SRO Justification: Technical Specification application of required actions in accordance with the rules of application requirements (section 1)

Monticello NRC Written Exam Page: 36 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 87 Technical

References:

T.S. 3.5.1, T.S. rules of application requirements section 1 Proposed references to be provided: T.S. Section 1.3, COMPLETION TIMES T.S. Section 3.0 T.S. Section 3.5.1, EMERGENCY CORE COOLING SYSTEM (ECCS)

Learning Objective: M-8107L-005 LO 9, 10b Cognitive Level: Higher X Lower Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 37 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 88 Points: 1.0 The plant is in MODE 3 making preparations for a reactor re-start. All required surveillances are complete for entry into Mode 2. Given the following events:

• During the plant shutdown, IRM Channel 15 was bypassed due to output failing to zero.

• Just moments ago, it was determined that the range switch for IRM Channel 17 was malfunctioning.

Choosing from the following list:

1. Repair/Replace the malfunctioning range switch

2. Place IRM Channel 17 in the trip condition

3. Manually insert a half scram on RPS Channel A

4. Manually insert a half scram on RPS Channel B What action(s) will allow the Mode switch to be placed in STARTUP/HOT STANDBY.

A. 1 ONLY B. 1 OR 2 ONLY C. 1 OR 2 OR 3 ONLY D. 1 OR 2 OR 4 ONLY Monticello NRC Written Exam Page: 38 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 88 Answer Explanation Answer: D DISTRACTOR ANALYSIS A. Is incorrect: will permit mode switch, BUT 2 and 4 are also allowed B. is incorrect: will permit mode switch, BUT 4 is also allowed.

C. is incorrect: 3 is not allowed because IRM 15 and IRM 17 are associated with RPS CH B D. is correct: There are 8 IRM channels, 4 in each RPS channel. Per TS 3.3.1.1, 3 IRM channel are required per channel, so given the above conditions that 2 channels within a RPS channel are INOP this LCO is NOT met, thus Condition A must be entered. A Range Switch is in each IRM channel and it allows the operator to follow power when in the IRM range. Switching the range switch upscale too early or too far will result in a downscale trip and rod block. Switching upscale too late or inadvertently switching downscale may result in Hi-Hi scram trip. With a faulty range switch, it must be addressed with actions other than a bypass (if the IRM is bypassed, the range switch must be ranged up to allow bypassing the SRM interlocks). If the Range switch is faulty, no matter high or low, it must be addressed before moving up in Mode. 1, 2, & 4 will all allow indefinite operation per LCO 3.0.4.a.

Question Background Tier: 2 Group: 1 K/A: (215003 IRM) A2.06 Ability to (a) predict the impacts of the following on the INTERMEDIATE RANGE MONITOR (IRM) SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 / 45.6) Faulty range switch Importance Rating: 3.2 10 CFR Part 55: CFR: 41.5 / 45.6 10 CFR 55.43.b: 2 (Facility operating limitations in the TS and their bases)

K/A Match: K/A is matched by testing the applicants ability to predict the impacts a faulty range switch will have on the IRM system and based on those impacts determine whether a mode switch is permitted in accordance with LCO 3.0.4 SRO Justification: Application of generic Limiting Condition for Operation (LCO) requirements (LCO 3.0.1 thru 3.0.7)

Monticello NRC Written Exam Page: 39 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 88 Technical

References:

TS 3.3.1.1, LCO 3.0.4a, and B.05.01.01-01, -02, and -05 Proposed references to be provided: TS 3.3.1.1 and LCO 3.0.4a Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: Question 87 of 2009 MNGP AUDIT Written Exam - slightly modified question stem to explicitly incorporate faulty range switch Comments: There are relatively easy modification that can be made to this question in order to declare it significantly modified (e.g., changing the IRM channels that are INOP, making bypass an option, etc.).

Monticello NRC Written Exam Page: 40 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 89 Points: 1.0 A plant startup is in progress. At 0400, the following conditions exist:

• RPV pressure being controlled at 150 psig on the MPR

• One Turbine Bypass Valve is 80% open to support HPCI, RCIC and SRV testing

• HPCI and SRV testing was completed satisfactorily, RCIC testing is in progress

• During RCIC testing, the maximum flow rate achieved by the RCIC pump was 350 gpm Select the statement below that correctly implements the requirements of Technical Specification LCO 3.5.3 (RCIC SYSTEM). Refer to the attached copy of TS LCO 3.5.3.

A. RPV pressure must be maintained 150 psig, and the Reactor Mode switch must be placed in SHUTDOWN before 1600.

B. The RCIC system must be capable of meeting the requirements of SR 3.5.3.3 before 1600.

Until then, plant heat-up and pressurization beyond 165 psig is prohibited.

C. Plant heat-up and pressurization to 920 psig may continue since the HPCI system testing was completed satisfactorily, but entry into MODE 1 is prohibited until the RCIC system is capable of meeting the requirements of SR 3.5.3.2.

D. Plant heat-up and pressurization to 920 psig may continue, and entry into MODE 1 is permitted, since the HPCI system testing was completed satisfactorily, but the RCIC system must be capable of meeting the requirements of SR 3.5.3.2 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of entering MODE 1.

Monticello NRC Written Exam Page: 41 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 89 ANSWER EXPLANATION Answer B DISTRACTOR ANALYSIS The flow tests for the RCIC System are performed at two different pressure ranges such that system capability to provide rated flow against a system head corresponding to reactor pressure is tested both at the higher and lower operating ranges of the system. Adequate reactor steam pressure must be available to perform these tests. Additionally, adequate steam flow must be passing through the turbine bypass valves to continue to control reactor pressure when the RCIC System diverts steam flow. Therefore, sufficient time is allowed after adequate pressure and flow are achieved to perform these SRs. Reactor steam pressure must be 950 psig to perform SR 3.5.3.2 and 150 psig to perform SR 3.5.3.3. Adequate steam flow is represented by at least one turbine bypass valve 80% open. The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for performing the flow test, after the required pressure and flow are reached, is sufficient to achieve stable conditions for testing and provides reasonable time to complete the SRs.

A. Incorrect; Startup procedure 2167 implies a limit of 150 psig, but SR 3.5.3.3 allows pressure to be raised up to 165 psig for performing the test. MODE 3 entry is not required unless RCIC operability cannot be restored within 14 days. Plausibe since TS LCO 3.5.3 is not met and Required Actions B.1 and B.2, while not applicable specify placing plant in MODE 3 and limiting RPV pressure to 150 psig.

B. Correct; SR 3.5.3.3 (and associated bases) specifies that the surveillance requirement must be met within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of achieving the conditions necessary to perform the test. With TS LCO 3.5.3 not being met, heat up and pressurization beyond 150 psig (except as permitted to perform SR 3.5.3.3) is not permitted until the test is completed satisfactorily.

C. Incorrect, see B. Plausible if applicant believes that successful completion of the surveillance test is not required. LCO 3.5.3 would not be met and there is no allowance to enter a higher mode of applicability (RPV pressure > 150 psig) without the LCO being met D. Incorrect, LCO 3.5.3 would not be met and there is no allowance to enter a higher mode of applicability (RPV pressure > 150 psig) without the LCO being met.

Monticello NRC Written Exam Page: 42 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 89 Question Background Tier: 2 Group: 1 K/A: 217000 RCIC A2.11 Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Inadequate system flow Importance Rating: 3.2 10 CFR Part 55: 41.5 10 CFR 55.43.b: 2 K/A Match: The applicant is given plant conditions which have RCIC in service to confirm continued operability, the system does not meet S.R. requirements for system flow.

IAW TSs the plant startup to 920 psig is not permitted.

SRO Justification: The questions requires the applicant to understand and apply T.S. S.R. 3.0.2 to determine the latest time the RCIC surveillance must be completed to meet its specified frequency, and if startup actions can continue with the surveillance not being met, LCO 3.0.4.b is not permitted by T.S. 3.5.3 Technical

References:

Technical Specifications 2167 [plant startup]

Proposed references to be provided: Technical Specifications Section 3.5.3 RCIC System Learning Objective: M-8107L-003: 8, 10(b)

Cognitive Level: Higher X Lower Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 43 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 90 Points: 1.0 The plant is beginning a refuel outage with the Mode Switch in SHUTDOWN, reactor coolant temperature is 160°F and several reactor vessel head closure bolts de-tensioned.

Additional conditions are as follows:

• CRDH is injecting to the RPV

• RWCU is in the Dump Mode providing RPV level control

• An OPDRV (Operation with Potential to Drain the Reactor Vessel) is in progress A LOCA from the RWCU system occurs resulting in RPV water level lowering to -58 inches; which of the following EAL classifications must be made?

A. ALERT IAW CA1.1 B. ALERT IAW CA2.1 C. Site Area Emergency IAW CS1.1 D. Site Area Emergency IAW CS2.1 Monticello NRC Written Exam Page: 44 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 90 ANSWER EXPLANATION Answer: B DISTRACTOR ANALYSIS The examinee must first determine from the initial stem conditions that the plant is in Mode 5. Once that is determined the correct call would be CA2.1 due to RPV water level lowering below -47 inches.

A is incorrect: This would be the correct call if the plant was in Mode 4.

C is incorrect: This would be correct if water level falls below -53 inches and secondary containment NOT established in Mode 4. With the OPDRV in progress, secondary containment would be required and the plant is in Mode 5.

D is incorrect: This would be correct if water level falls below -53 inches and secondary containment NOT established in Mode 5. With the OPDRV in progress, secondary containment would be required Monticello NRC Written Exam Page: 45 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 90 QUESTION BACKGROUND Tier: 2 Group: 1 K/A: (259002 Reactor Water Level Control) 2.1.20 Ability to interpret and execute procedure steps. (CFR: 41.10 / 43.5 / 45.12)

Importance Rating: 4.6 10 CFR Part 55: CFR: 41.10 / 43.5 / 45.12 10 CFR 55.43.b: 5 K/A Match: K/A is matched by testing the applicants ability to interpret water level on the EAL Matrix procedure and execute the EAL declaration appropriately for a loss of level control..

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

Technical

References:

EAL Matrix Procedure, Tech Spec Mode Table Proposed references to be provided: EAL Matrix Learning Objective: MT-OPS-BEP-001L LO 2 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: 2015 ILT Audit Exam & 2010 ILT NRC Exam Comments:

Monticello NRC Written Exam Page: 46 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 91 Points: 1.0 The plant was at rated conditions with the #12 RHR pump operating in Torus Cooling Mode when a transient occurred resulting in the following indications:

Time RPV Level RPV Pressure Drywell pressure Torus Temperature 1004 -50 inches 1020 psig 1.0 psig 75°F 1021 -55 inches 800 psig 1.8 psig 78°F 1023 -60 inches 600 psig 1.9 psig 83°F 1025 -70 inches 700 psig 2.0 psig 90°F ADS actuation was inhibited.

Based on the above conditions, complete the statements below:

At 1021 the #12 RHR pump __(1)__ be operating in the Torus Cooling Mode.

At 1025 you should direct Control Room personnel to __(2)__.

A. (1) will (2) Prevent ALL injection from CS and LPCI B. (1) will (2) Transfer A and B RHR from LPCI to Torus Cooling C. (1) will NOT (2) Prevent ALL injection from CS and LPCI D. (1) will NOT (2) Transfer A and B RHR from LPCI to Torus Cooling Monticello NRC Written Exam Page: 47 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 91 ANSWER EXPLANATION Answer D DISTRACTOR ANALYSIS A LPCI Initiation Signal on Low Low reactor water level for 15 minutes becomes active at 1019, this sends a signal to close MO-2007 and MO-2009 [torus cooling valves] so a path for torus cooling is no longer aligned. After the ECCS actuation, Torus temperature begins to rise at an increasing rate and at 1025 Torus Temperature has reached a value requiring use of all available Torus Cooling. EOP C.5-1100 also directs that CS and LPCI injection not needed for core cooling be prevented. The examinee must balance the need for increased injection and the need for Torus Cooling.

A. (1) Incorrect, MO- 2007 and 2009 received a closed signal at 1019, the flow path for torus cooling is no longer aligned (2) Incorrect, with RPV level continuing to decrease additional injection is needed; preventing injection from all low pressure ECCS sources would not be appropriate.

B. (1) Incorrect, MO- 2007 and 2009 received a closed signal at 1019, the flow path for torus cooling is no longer aligned (2) Correct; see above.

C. (1) Correct (2) Incorrect, with RPV level continuing to decrease additional injection is needed; preventing injection from all low pressure ECCS sources would not be appropriate.

D. (1) Correct (2) Correct; see above.

Monticello NRC Written Exam Page: 48 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 91 Question Background Tier: 2 Group: 2 K/A: 219000 RHR/LPCI: Torus/Pool Cooling Mode A2.14 Ability to (a) predict the impacts of the following on the RHR/LPCI:

TORUS/SUPPRESSION POOL COOLING MODE ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Loss of coolant accident Importance Rating: 4.3 10 CFR Part 55: 41.5 10 CFR 55.43.b: 5 K/A Match: The applicant is given a set of plant conditions that indicate a LOCA is occurring, the plant will reconfigure with the LOCA signal, the applicant must evaluate the effect of this on torus cooling and then mitigate the accident conditions by evaluating plant Parameters and selecting the appropriate procedure.

SRO Justification: see screening for SRO only linked to 10CFR55.43(b)(5)

Technical

References:

C.5-1100, C.5-1200, B.03.04-05, B.03.04-02 Proposed references to be provided: C.5-1100 & C.5-1200 Learning Objective: M-8107L-023, LO 8 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 49 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 92 Points: 1.0 Shortly after completing a full core off-load, an event occurred resulting in a leak in the reactor cavity bellows seals.

• SFP Level is currently 36 feet.

• SFP Level is dropping at 9 inches/hour

• Skimmer Surge Tank levels are 2 feet and stable

• Fuel Pool to Reactor Cavity gates have been installed Which of the following EOP actions is preferred to mitigate the above conditions?

A. Establish makeup with Condensate Service Water using the Filter/Demin Backwash connection; then restore the FPC&C System to operation.

B. Establish makeup with RHRSW via the Residual Heat Removal System, then lineup and operate RHR in the Fuel Pool Cooling Mode.

C. Restore the FPC&C System to operation, then establish makeup with Condensate Service Water using the Filter/Demin Backwash connection.

D. Lineup and operate RHR in the Fuel Pool Cooling Mode, then establish makeup with RHRSW via the Residual Heat Removal System.

Monticello NRC Written Exam Page: 50 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 92 ANSWER EXPLANATION Answer: A DISTRACTOR ANALYSIS A. CORRECT: Per AOP C.4-B.02-01.A the bottom of the weir is 373 at which point no water will be returning to the surge tanks. The SFP low level alarm (4-B-33) occurs at 367 and a short time later the Fuel Pooling Cooling pump(s) will trip on low surge tank level or low suction pressure. Pumps will not be able to be restarted until after SFP is restored to the point that water overflows to the surge tanks. Makeup using the Filter/Demineralizer Backwash connection preferred over RHRSW via RHR since the makeup source is the CST rather than river water. Per B.02.01-05, at >24.75 ft (SFP level > 10 ft above fuel racks) adequate radiation shielding is available. Thus, current conditions show adequate radiation shielding to the spent fuel assembly is present and access to the refuel floor is still available. EOP C.5-1300 entry is required below 367 SFP level and if level cannot be restored and maintained above 367 alternate SFP makeup shall be used, such as Condensate Service Water hose connection and RHRSW through RHR. The preferred normal makeup method (CST-32) is not available but the alternate method using the Filter/Demineralizer Backwash connection is available.

B, C, and D are incorrect but plausible variations of ARP, AOP and EOP mitigating strategies of a low SFP level.

B. Incorrect; makeup source is the river rather than the CST. Operating RHR in the Fuel Pool Cooling mode would be acceptable if Fuel Pool Cooling and Cleanup could not be restored.

C. Incorrect; SFP level must be restored before restoring Fuel Pool Cooling and Cleanup system to operation.

D. Incorrect; SFP level must be restored before operating RHR in the Fuel Pool Cooling Mode, and the CST is the preferred makeup source.

QUESTION BACKGROUND Tier: 2 Group: 2 K/A: (233000 Fuel Pool Cooling/Cleanup) A2.02 Ability to (a) predict the impacts of the following on the FUEL POOL COOLING AND CLEAN-UP ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: (CFR: 41.5 /

45.6) Low pool level Monticello NRC Written Exam Page: 51 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 92 Importance Rating: 3.3 10 CFR Part 55: CFR: 41.5 / 45.6 10 CFR 55.43.b: 4 and 5 K/A Match: K/A is matched by testing the applicants ability to predict the impacts a low SFP level will have on FPCC system and based on those impacts what procedures or operations should be used to mitigate those consequences. Specifically, the question is testing the applicants understanding of the impacts a low SFP level will have to potential operations and the ability to determine what procedural steps are required to address the low SFP level.

SRO Justification: Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions; Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations Technical

References:

Ops Man B.02.01-03 and -05, ARP 4-B-33, C.4-B.02.01.A, C.5-1300, and M-8107L-022 (Fuel Pool Cooling And Cleanup) Rev. 14 Proposed references to be provided: C.5-1300 Learning Objective: 9 of M-8107L-022 (Fuel Pool Cooling And Cleanup) Rev. 14 Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 52 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 93 Points: 1.0 A plant startup is in progress with the following conditions:

• RHR A is in Shutdown Cooling

• Reactor coolant temperature is 120°F

• The Mode Switch is in shutdown The status of the RCS Leakage Detection Instrumentation is as follows:

• Drywell Floor Drain Sump Monitoring System is INOPERABLE

• Drywell Equipment Drain Sump Monitoring System is OPERABLE

• Drywell Floor Drain Sump is NOT overflowing to the Drywell Equipment Drain Sump Given the above information, can the unit be taken to MODE 2; why or why not? (T.S. 3.4.5 and LCO 3.0 are provided)

A. TS LCO 3.4.5 is NOT met and MODE 2 may NOT be entered.

B. TS LCO 3.4.5 is met and MODE 2 may be entered with NO additional requirements.

C. TS LCO 3.4.5 is NOT met, but MODE 2 may be entered because of a previously NRC approved exception as specified by applicability of TS LCO 3.0.4.c.

D. TS LCO 3.4.5 is NOT met, but MODE 2 may be entered because the associated ACTIONS to be entered permit continued operation for an unlimited period of time.

Monticello NRC Written Exam Page: 53 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 93 ANSWER EXPLANATION Answer C DISTRACTOR ANALYSIS A. Incorrect, LCO 3.0.4.c is applicable, so entry is permitted into the mode of applicability.

B. Incorrect, plausible distractor if the applicant does not evaluate the information in the stem correctly.

C. Correct, entry into mode 2 is permitted because there is an allowance stated in the specification that LCO 3.0.4.c is applicable.

D. Incorrect, TS 3.4.5 Action A has a 30 day time clock; therefore LCO 3.0.4.a is not applicable.

QUESTION BACKGROUND Tier: 2 Group: 2 K/A: 268000 Radwaste 2.2.38 Knowledge of conditions and limitations in the facility license.

Importance Rating: 4.5 10 CFR Part 55: 55.41.7 10 CFR 55.43.b: 1, 2 K/A Match: KA is matched by having the applicant evaluate the operability of a portion of the liquid radwaste system and determine if a mode change is permitted in accordance with the Technical Specifications.

SRO Justification: This question can only be answered by applying generic LCO requirements.

Technical

References:

Technical Specifications Proposed references to be provided: T.S. 3.4.5 and LCO 3.0 Monticello NRC Written Exam Page: 54 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 93 Learning Objective: M-8107L-101 9, 10b Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History: Significant modification from MNGP bank question 1391 Comments:

Monticello NRC Written Exam Page: 55 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 94 Points: 1.0 The plant is operating at 50% power. At 1100 you are performing your Shift Supervision completion review of a surveillance that documents the AS LEFT trip settings for the Turbine Control Valve Fast Closure instruments. Trip settings are as follows:

• PS-7110: 166.4 psig

• PS-7111: 167.4 psig

• PS-7112: 168.7 psig

• PS-7113: 167.9 psig Given NX-7834-67-7 & 8; which one of the following actions is required, and if performed, will prevent a Technical Specification violation?

A. Restore RPS trip capability by 1200.

B. Place either RPS channel in trip by 1700.

C. Place RPS Channel A in trip by 2300.

D. Place RPS Channel B in trip by 2300.

Monticello NRC Written Exam Page: 56 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 94 Answer Explanation Answer: A DISTRACTOR ANALYSIS RPS A (PS-7112 & PS-7113); RPS B (PS-7110 & PS-7111). All pressure switches are required to be operable ( 167.8 psig) when 45% RTP. With the given conditions both channels in RPS B trip system are inoperable which makes the function lost. This will require entry into TS 3.3.1.1 conditions A and C. Choice A represents condition C and its required action. If this action is not taken first by 1200 compliance with TS will not be maintained.

B is incorrect: Plausible action to take if the switches are not recognized to both be on RPS B C is incorrect: Plausible action to take if candidate does not recognize the loss of RPS trip capability and incorrectly identifies A and B pressure switches.

D is incorrect: This action is required but compliance will not be maintained if this only was performed because RPS trip capability is lost.

Question Background Tier: 3 Group: 1 K/A: 2.1.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation. (CFR:

41.5 / 43.5 / 45.12 / 45.13)

Importance Rating: 4.7 10 CFR Part 55: 41.5 / 43.5 / 45.12 / 45.13 10 CFR 55.43.b: 5 K/A Match: K/A is matched by testing the applicants ability to evaluate the plant conditions (Reactor Power) and interpret instrument settings.

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

Monticello NRC Written Exam Page: 57 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 94 Technical

References:

TS 3.3.1.1, NX-7834-67-7 & 8 Proposed references to be provided: TS 3.3.1.1, NX-7834-67-7 & 8 Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: Question 87 of 2009 MGNP ILT Comments: Stem slightly modified lower reactor power closer to the operability requirement.

Monticello NRC Written Exam Page: 58 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 95 Points: 1.0 The plant was at rated conditions when Drywell pressure begins to slowly rise. The operators are executing C.4-B.04.01.F (LEAK INSIDE PRIMARY CONTAINMENT). The CRS has briefed the crew that a manual Reactor Scram will be inserted when Drywell pressure reaches 1.6 psig.

Using FP-PA-HU-05 (DECISION MAKING) Attachment 5, determine what level of decision making the CRS has employed for this situation?

A. Tier 1 B. Tier 2 C. Type 1 ODMI D. Type 2 ODMI Monticello NRC Written Exam Page: 59 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 95 ANSWER EXPLANATION Correct Answer: B DISTRACTOR ANALYSIS JUSTIFICATION: An example of Tier 2 Decision Making is Establishing conservative values for operation, at what point action will be taken prior to reaching an automatic set point. In this case the automatic setpoint would be 1.84 psig which will initiate an automatic reactor scram.

A, C and D are incorrect: Plausible decision making levels found in Attachment 3.

QUESTION BACKGROUND TIER: 3 GROUP: 2.1 K/A: 2.1.39 Knowledge of conservative decision making practices.

Importance Rating: SRO 4.3 10 CFR: 55.43b(1)

SRO ONLY JUSTIFICATION: SRO Only task for Conservative Decision Making

REFERENCE:

FP-PA-HU-05 (DECISION MAKING)

REFERENCE PROVIDED DURING EXAM: FP-PA-HU-05 (DECISION MAKING) Attachment 5 OBJECTIVE: M8108L-038, 2 COG LEVEL: 3 SPR QUESTION SOURCE: BANK -- 2015 MNGP ILT NRC Written Exam DIFFICULTY: 2 Comments:

Monticello NRC Written Exam Page: 60 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 96 Points: 1.0 Which of the following conditions would SOLELY render a supported system no longer capable of performing its intended safety function? Refer to the provided Section 3.0 of MNGP Technical Specifications.

A. One or more snubbers are not capable of providing their associated support function(s), to more than one subsystem of a multiple subsystem supported system, for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

B. One or more snubbers are not capable of providing their associated support function(s), to a single subsystem of a multiple subsystem supported system, or to a single subsystem supported system, for 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

C. One or more barriers, required to support operability of a supported system, are not capable of performing their related support function(s) for 14 days.

D. A support system is inoperable, and a required system - redundant to the system(s) supported by the inoperable support system - is also inoperable.

Monticello NRC Written Exam Page: 61 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 96 Answer Explanation Answer: D DISTRACTOR ANALYSIS A and B are incorrect: the times stated are NOT long enough for the supported system to be declared INOP; LCO 3.0.8 states that the supported system is considered to remain capable of performing its intended safety when one or more snubbers are not capable of performing their associated supported function(s) for a specific allowed time - 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> for a multiple subsystem supported system and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> for a single subsystem supported system or to a single subsystem supported system.

C is incorrect: the time stated is NOT long enough for the supported system to be declared not capable of performing its safety function; LCO 3.0.9 states that the supported system is not considered INOP solely due to required barriers not being able to perform their related support function(s). 30 days is allowed before declaring the supported system(s) incapable of performing the safety function.

D is correct: According to LCO 3.0.6, and TS 5.5.10, a loss of safety function may exist when a support system is inoperable, and a required system redundant to system(s) supported by the inoperable support system is also inoperable.

For example, based on the Tech Spec BASES example (EXAMPLE B 3.0.6-1), considering the figure below: If System 2 of Division A is inoperable and System 5 of Division B is inoperable, a loss of safety function exists in supported System 5.

Monticello NRC Written Exam Page: 62 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 96 Question Background Tier: 3 Group: 2 K/A: 2.2.37 Ability to determine operability and/or availability of safety related equipment.

(CFR: 41.7 / 43.5 / 45.12)

Importance Rating: 4.6 10 CFR Part 55: 41.7 / 43.5 / 45.12 10 CFR 55.43.b: 5 K/A Match: K/A is matched by testing the applicants ability to determine the operability of a safety related equipment by using the motherhood LCOs.

SRO Justification: Application of generic LCO requirements Technical

References:

OWI-03.05 Rev. 5, LCO 3.0.6, 3.0.8, and 3.0.9, TS 5.5.10 Proposed references to be provided: TS Section 5.5.10, TS Section 3.0 Learning Objective:

Cognitive Level: Higher X Lower _

Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 63 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 97 Points: 1.0 An EOP Flowchart change to C.5-1200 is being developed IAW 4-AWI-08.16.01 (MONTICELLO EOP AND BEYOND-DESIGN-BASIS GUIDELINE MAINTENANCE PROGRAM) due to a recent change in the BWROG EPGs.

IAW 4-AWI-08.16.01, complete the statements below?

(1) The performance of a 10CFR50.59 __(1)__ is required for all EOP changes.

(2) EOP Verification and Validation is required to be performed __(2)__ PORC review.

A. (1) screening (2) before B. (1) screening (2) after C. (1) evaluation (2) before D. (1) evaluation (2) after Monticello NRC Written Exam Page: 64 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 97 Answer Explanation Answer A DISTRACTOR ANALYSIS 4-AWI-08.16.01 requires a 10CFR50.59 screening to be performed for all EOP changes and verification and validation SHALL be performed prior to submitting the flowchart revision to PORC for review.

A. Correct B. (1) Correct, (2) Incorrect, the verification and validation is required prior to PORC review to ensure technical accuracy and usability.

C. (1) Incorrect, screening is required by procedure (2) Correct D. (1) Incorrect, screening is required by procedure (2) Incorrect, the verification and validation is required prior to PORC review to ensure technical accuracy and usability.

Question Background Tier: 3 Group: N/A K/A: Equipment control 2.2.6 Knowledge of the process for making changes to procedures Importance Rating: 3.6 10 CFR Part 55: 41.10 10 CFR 55.43.b: 3 K/A Match: The applicant is asked to evaluate requirements to make changes to the EOPS IAW station procedures and to ensure regulatory requirements are met Monticello NRC Written Exam Page: 65 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 97 SRO Justification: Knowledge of the process for changing plant procedures Technical

References:

4-AWI-08.16.01 Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History:

Comments:

Monticello NRC Written Exam Page: 66 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 98 Points: 1.0 The plant was at rated conditions when an event occurred. Twenty (20) minutes later the following conditions exist:

• A General Emergency has been declared

• The TSC and EOF have NOT yet been manned

• 5790-401-01 (EMERGENCY EXPOSURE AUTHORIZATION FORM) has been initiated for Joe Radworker and is ready for your review

• Joe is assigned to manipulate several valves for protection of valuable property

• Radiation Protection estimates that Joe will receive the following exposure:

o Whole Body: 7 REM o Lens of the Eye: 25 REM o Hands and Forearms: 75 REM

• A.2-401 Figure 7.1 (EMERGENCY WORKER EXPOSURE LIMITS) is provided on the following page.

As the Emergency Director; which one of the following is correct, and why?

A. Approve the 5790-401-01 as the estimated exposures are within limits.

B. Do NOT approve the 5790-401-01 as the estimated exposure to the Whole Body exceeds the limit.

C. Do NOT approve the 5790-401-01 as the estimated exposure to the Hands and Forearms exceeds the limit.

D. Do NOT approve the 5790-401-01 as the estimated exposure to the Lens of the Eye exceeds the limit.

Monticello NRC Written Exam Page: 67 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Monticello NRC Written Exam Page: 68 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 98 Answer Explanation Answer: A DISTRACTOR ANALYSIS The allowed Emergency Worker Exposure Limits for an emergency activity to protect valuable equipment is 10 REM. NOTE 3 on Figure 7.1 allows 3 times this number for lens of the eye and 10 times this number for the skin and/or extremities. Based on that the limits are as follows:

Lens of the eye: 30 REM Hands and forearms: 100 REM Whole Body: 10 REM B. C and D are incorrect: Plausible for incorrect application of Figure 7.1.

Question Background Tier: 3 Group: 3 K/A: 2.3.4 Knowledge of radiation exposure limits under normal or emergency conditions.

(CFR: 41.12 / 43.4 / 45.10)

Importance Rating: 3.7 10 CFR Part 55: 41.12 / 43.4 / 45.10 10 CFR 55.43.b: 4 K/A Match: K/A is matched by testing the applicants knowledge of the radiation exposure limits under emergency conditions SRO Justification: Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative procedures.

Technical

References:

A.2-401 Proposed references to be provided: A.2-401 Figure 7.1 (inserted within question)

Learning Objective:

Monticello NRC Written Exam Page: 69 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 98 Cognitive Level: Higher X Lower _

Question Source: New _

Modified Bank _

Bank X Question History: Question 98 from 2015 MNGP Audit Exam Comments:

Monticello NRC Written Exam Page: 70 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 99 Points: 1.0 The reactor is at rated conditions with the HPCI QUARTERLY PUMP FLOW AND VALVE TEST (0255-06-IA-1) in progress as scheduled. During the test, average Torus water temperature reached 91°F.

Complete the following statement describing the correct action by the CRS as it pertains to the EOPs?

C.5-1200 (PRIMARY CONTAINMENT CONTROL)

A. IS NOT required to be entered as the limit is 100°F during testing that adds heat to the Torus.

B. IS NOT required to be entered as the temperature rise is directly attributable to a surveillance test.

C. IS required to be entered, but may be subsequently exited when it is determined that an emergency condition does not exist.

D. IS required to be entered, and may be exited ONLY when Torus temperature has been lowered to less than 90°F with RHR in Torus Cooling.

Monticello NRC Written Exam Page: 71 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 99 ANSWER EXPLANATION Answer: C DISTRACTOR ANALYSIS JUSTIFICATION: EOPs must be entered any time the plant is not in cold shutdown or outage conditions and an entry condition occurs, even if the condition exists for only a short time and then clears. If an EOP entry was because of a pre-planned evolution, the EOPs may be exited when it is determined that an emergency does not exist.

A is incorrect: 100°F is the LCO Action entry requirement for Torus temperature with testing in progress, but entry into C.5-1200 is still required > 90°F.

B is incorrect: EOP entry is required even for pre-planned evolutions such as surveillance testing.

D is incorrect: EOPs may be exited if no emergency condition exists for pre-planned evolutions.

QUESTION BACKGROUND TIER: 3 GROUP: 4 CATEGORY: Emergency Procedures / Plan K/A: 2.4.14 Knowledge of general guidelines for EOP usage.

IMPORTANCE: SRO 4.5 10 CFR: 55.43b(1)

SRO ONLY JUSTIFICATION: EOP exit requirements

REFERENCE:

C.5.1-1000 REFERENCE PROVIDED DURING EXAM: None LESSON PL: MT-ILT-EOP-001L OBJECTIVE: 4, 6 COG LEVEL: 1 P Question Source: ILT Bank (2015 NRC Audit)

Comments:

Monticello NRC Written Exam Page: 72 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question 100 Points: 1.0 Which of the following is an INITIAL Protective Action Recommendation (PAR) for a General Emergency Classification made in the Control Room NOT due to a loss of the containment barrier?

A. Evacuate a 2 mile radius and 2-5 mile down wind. All others monitor and prepare.

B. Evacuate a 2 mile radius and 2-10 mile down wind. All others monitor and prepare.

C. Contact the State Planning Chief or State Duty Officer if State EOC is not activated.

D. No protective action recommendation is appropriate when projected plume dose rates do NOT exceed 1000 mrem (TEDE) OR 5000 mrem (CDE) thyroid dose.

Monticello NRC Written Exam Page: 73 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 100 Answer Explanation Answer: A DISTRACTOR ANALYSIS A. Correct: According to A.2-204, declaration of a General Emergency requires timely initial protective action recommendations (PARs) to off-site agencies. Under these circumstances, NO dose projections are required for formulating the initial off-site protection action recommendation UNLESS there is a Rapidly Progressing Severe Accident. A Rapidly Progressing Severe Accident is a General Emergency (GE) with rapid loss of containment integrity (emergency action levels indicate containment barrier loss) and loss of ability to cool the core. Thus, since this GE is NOT due to loss of containment barrier per the EAL, the applicant shall conclude a GE exists WITHOUT a Rapidly Progressing Severe Accident. Figure 7.3.A of A.2-204 is used. The applicant should conclude the initial PAR is to Evacuate a 2 mile radius and 2-5 miles downwind. All others monitor and prepare.

B. Incorrect: 2-10 miles downwind is not the initial PAR distance if a Rapidly Progressing Severe Accident does NOT exist (there is not a loss of containment barrier)

C. Incorrect: this action is taken AFTER the recommendation is made D. Incorrect: Initial PARs for a General Emergency involving loss of physical control or core damage require immediate evacuation of the general public and are justified without dose projection; recommending off-site protective actions shall not be delayed.

Question Background Tier: 3 Group: 4 K/A: 2.4.44 Knowledge of emergency plan protective action recommendations. (CFR:

41.10 / 41.12 / 43.5 / 45.11)

Importance Rating: 4.4 10 CFR Part 55: CFR: 41.10 / 41.12 / 43.5 / 45.11 10 CFR 55.43.b: 5 K/A Match: K/A is matched by testing the applicants knowledge of the emergency plan protective action recommendations, specifically after a General Emergency Classification is made. The applicants knowledge of PARs and the E-Plan are tested.

Monticello NRC Written Exam Page: 74 of 75 2016

FINAL EXAMINATION ANSWER KEY ILT NRC SRO WRITTEN EXAM Question Number: 100 SRO Justification: E-Plan Technical

References:

A.2-204, E-Plan Proposed references to be provided: None Learning Objective:

Cognitive Level: Higher _

Lower X Question Source: New X Modified Bank _

Bank _

Question History: Question 100 of 2007 MNGP ILT NRC - significantly modified (New) by replacing distractor C and including more info in the question stem Comments:

Monticello NRC Written Exam Page: 75 of 75 2016

2016 MONTICELLO WRITTEN EXAMS PROVIDED REFERENCES RO EXAM

1. Drawings NX-7834-67-9 & 10 [Q4]

2. ARP 8-A-34 (1-BUS 345 KV LOW FREQ) [Q20]

3. Figure 2 of Ops Man B.09.02-02 [Q20]

SRO EXAM

1. Form 5790-101-02, MONTICELLO NUCLEAR GENERATING PLANT EMERGENCY ACTION LEVEL MATRIX [Q77; Q82; Q90]

2. Technical Specifications

a. TS Section 1.3, Completion Times [Q87]

b. TS Section 3.0, LCO Applicability [Q78; Q87; Q88; Q96]

c. TS Section 3.0 SR Applicability [Q86]

d. TS Section 3.3.1.1 [Q88; Q94]

e. TS Section 3.4.5 [Q93]

f. TS Section 3.4.7 [Q78]

g. TS Section 3.4.8 [Q78]

h. TS Section 3.4.9 [Q78]

i. TS Section 3.5.1 [Q87]

j. TS Section 3.5.3 [Q89]

k. TS Section 3.8.1 [Q78]

l. TS Section 3.8.4 [Q78]

m. TS Section 3.8.7 [Q78]

n. TS Section 3.8.8 [Q78]

o. TS Section 5.5.10 [Q96]

3. Steam Tables [Q78]

4. C.5-1100 with entry conditions blanked out. [Q79; Q91]

5. C.5-1200 with entry conditions blanked out [Q79; Q80; Q85; Q91]

6. C.5-1300 with entry conditions blanked out [Q92]

7. Drawings NX-7834-67-7 & 8 [Q94]

8. FP-PA-HU-05 (DECISION MAKING) Attachment 5 [Q95]