ML15009A287

2014-12 Draft Written Exam

ML15009A287
Person / Time
Site:	Callaway
Issue date:	12/17/2014
From:	Vincent Gaddy Operations Branch IV
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Download: ML15009A287 (217)

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NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 059 Main Feedwater Group # 1 K/A # A3.04 Importance Rating 2.5 Ability to monitor automatic operation of the MFW, including: Turbine driven feed pump.

Question #1 Given the following plant conditions:

• A plant shutdown is in progress

• Reactor power is currently 25%

• Feed Pump Master Control DELTA P setpoint is currently 71 psid

• The transfer from MFRVs to MFRV Bypass Valves is being performed using the Automatic Method What is the expected response of the Feed Pump Master Control DELTA P setpoint?

A. Lowers to 45 psid B. Stays at 71 psid C. Raises to 149 psid D. Raises to 215 psid Answer: D Explanation: Per OTN-AE-00001 when the automatic method of transfer from the MFRV the MFRV Bypass valves is used the Feed Pump Master Control DELTA P setpoint automatically raises to 215 psid.

A. Incorrect - This is the minimum value for the Feed Pump Master Control DELTA P setpoint.

This is a plausible distractor if the student incorrectly assumes the system goes to minimum value when transfer to the MFRV bypass occurs B. Incorrect - This is the current program value. This is a plausible distractor if the student incorrectly assumes the system maintains the current value when transfer to the MFRV bypass occurs C. Incorrect - This is the maximum value for the Feed Pump Master Control DELTA P setpoint when the MFRV is in use. This is a plausible distractor if the student incorrectly assumes the system goes to maximum value for the MFRV when transfer to the MFRV bypass occurs D. Correct - see explanation above

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): OTN-AE-00001, Feedwater System, Rev 52 References to be provided to applicants during examination: None Learning Objective: Main Feedwater System - AE, Obj E, DESCRIBE the operation, including signal inputs, of the MFW pump speed control system and EXPLAIN the control response to input failures.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 061 Auxiliary/Emergency Feedwater Group # 1 K/A # 2.2.38 Importance Rating 3.6 Knowledge of conditions and limitations in the facility license.

Question #2 Given the following plant conditions:

• At 1100 a plant cooldown at 50 F/Hr from 549°F is initiated using the AUX FEED system and S/G PORVs

• CST level is at 89% (418,000 gal)

What time would T.S. entry be made, if any, if AFW flow is maintained constant at 385,000 lbm/hr? (8.345 lbm/gal)

A. Never, no T.S. entry is required B. 1130 C. 1400 D. 1500 Answer: C Explanation: The TS for the CST is 281,000 gallons. The TS for the CST is applicable in MODES 1, 2, and 3. Mode 4 will be reached when the Reactor temp is less than 350°F. The TS for the RWST is 394,000 gallons.

Cooldown: 549°F - 350°F = 199°F. 199°F / 50°F/hour = slightly less than 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. Mode 4 will be reached at 1500.

Feedwater conversion: 385,000 lbm/hr / 8.345 lbm/gal = 46,135 gal/hr CST level calculation: 418,000 - 281,000 = 137,000 gallons. 137,000 gal / 46,135 gal/hr= 2.97 hours0.00112 days <br />0.0269 hours <br />1.603836e-4 weeks <br />3.69085e-5 months <br />.

RWST level calculation: 418,000 - 394,000 = 24,000 gallons. 24,000 gal / 46,135 gal/hr = 30 minutes

NRC Site-Specific Written Examination Callaway Plant Reactor Operator A. Incorrect - Based on the above calculations the CST level TS is met prior to entry into Mode

4. IF the candidate incorrectly calculates the cooldown rate or the level change in the CST, the candidate could incorrectly assume that Mode 4 is met before the CST level is met and would conclude that TS entry is not required due to not being applicable in Mode 4.

B. Incorrect - Based on the above calculations, if the students uses the RWST low level TS, (RWST and CST are both large tanks with similar TS low level values) this is when the TS limit would be reached.

C. Correct - Based on the above calculations, this is when the CST would reach the TS level of 281,000 gal. in Mode 3.

D. Incorrect - Based on the calculations above this is when the cooldown rate would result in entry into Mode 4 where the CST T/S is no longer applicable.

Technical Reference(s): Tech Spec Table 1.1-1 and TS 3.7 References to be provided to applicants during examination: None Learning Objective: F. DEMONSTRATE the proper use of Technical Specifications, Bases and FSAR Chapter 16.

Question Source: Bank # ______

Modified Bank # __L5123____

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

TABLE 1.1-1 (PAGE 1 OF 1) MODES MODE TITLE REACTIVIT  % RATED THERMAL AVERAGE REACTOR Y POWER(a) COOLANT CONDITION TEMPERATURE (° F)

(keff)

Power 1

Operation 0.99 >5 NA 2 Startup 0.99 5 NA

NRC Site-Specific Written Examination Callaway Plant Reactor Operator 3 Hot Standby < 0.99 NA 350 Hot 4

Shutdown(b) < 0.99 NA 350 > Tavg > 200 Cold 5

Shutdown(b) < 0.99 NA 200 6 Refueling(c) NA NA NA (a) Excluding decay heat.

(b) At least 53 of 54 reactor vessel head closure bolts fully tensioned.

(c) Two or more reactor vessel head closure bolts less than fully tensioned.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 061 Auxiliary/Emergency Feedwater Group # 1 K/A # K1.01 Importance Rating 4.1 Knowledge of the physical connections and/or cause effect relationships between the AFW and the following systems: S/G system.

Question #3 Given the following plant conditions:

• A plant trip occurs from 100% power

• A lockout of ESF Train 'A' 4160 Volt Bus NB01 occurs

• The Turbine Driven Auxiliary Feedwater Pump fails to start Which Steam Generators (SGs) are being supplied by Auxiliary Feedwater?

A. 'A' and 'B' B. 'A' and 'D' C. 'B' and 'C' D. 'C' and 'D' Answer: B Explanation:

A. Incorrect. See B.

B. Correct. Upon receipt of a reactor trip, the subsequent shrink in the steam generators combined with the loss of the normal feedwater from the Turbine driven feedwater pumps will cause steam generator levels to lower to 17%, at which time the AFW system receives an actuation signal to start. Motor Driven AFW Pump A is designed to feed only Steam Generators B & C, Motor Driven AFW Pump B is designed to feed only Steam Generators A & D. The Turbine Driven AFW Pump can feed All Four Steam Generators. Given the Conditions where the TDAFW pump and the Power Supply to A MDAFW pump are lost, this leaves only B MDAFW pump to supply A & D Steam Generators.

C. Incorrect. See B.

D. Incorrect. See B.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): Drawing M-22AL01 REV 43.

References to be provided to applicants during examination: None.

Learning Objective: T-61.0110 LP-25 Obj. C. Draw, Label, and Explain a one line diagram of the AFW system to include MFAFW Pumps, TDAFW Pumps, AFW Flow Control Valves, showing the system alignment under Normal Conditions and Aux Feedwater Actuation Signal, or SI present.

Question Source: Bank # ______

Modified Bank # ___L16582___

New _______

Question History: Last NRC Exam ______Callaway 2013 ILT NRC EXAM______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.4)

Comments:

This question has been modified from a question used on the 2013 ILT NRC Exam.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 062 AC Electrical Distribution Group # 1 K/A # K4.03 Importance Rating 2.8 Knowledge of ac distribution system design feature(s) and/or interlock(s) which provide for the following: Interlocks between automatic bus transfer and breakers.

Question #4 Which of the following malfunctions will result in an AUTOMATIC closure of PB0401, PB04 to PB03 Crosstie breaker?

A. 186 Lockout on PB03 Bus Feeder Breaker (PB0306)

B. Manually opening PB04 Bus Feeder Breaker (PB0402)

C. 1 of 2 Pressure Fault switches on XPB03 Transformer actuates D. Differential Current Lockout of XPB04 Transformer Feeder Breaker (PA0208)

Answer: D Explanation:

A. Incorrect. Plausible because a lockout on the XPB03 feeder breaker from the 13.8 kV bus will result in an automatic closure of the crosstie breaker, however the XPB03 to PB03 feeder breaker(PB0306) does not tie into the circuit for automatic closure of the crosstie breaker.

B. Incorrect. Plausible because manual opening of the breaker causes an undervoltage conditions resulting in automatic actuations on other busses (ie. NB01 or NB02). Incorrect because even though the cross tie breaker PB0401 can be closed manually if the fault does not exist on PB04, this will not result in an Automatic closure of the crosstie breaker.

C. Incorrect. Plausible because an over pressure fault on either the XPB03 or XPB04 transformers will result in an Automatic closure of the crosstie breaker. However this actuation requires 2/2 pressure switches to actuate to cause the transfer. With only 1 of 2 switches actuating it will only result in an alarm in the control room.

D. Correct. With the system in a normal lineup, a Differential Overcurrent condition will result in a Lockout of the PA0208 Breaker. This will automatically trip the PB04 Bus Feeder Breaker PB0402, and close the PB0401 (PB04 to PB03 cross tie breaker).

Technical Reference(s): Drawing E-23PB14, PB03 & PB04 Bus Tie Breaker References to be provided to applicants during examination: None.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Learning Objective: T61.0110 LP-02, Obj. E. Discuss the conditions that will trip the following breakers: 1. Station Service Transformer Feeder Breakers.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 062 AC Electrical Distribution Group # 1 K/A # K2.01 Importance Rating 3.3 Knowledge of bus power supplies to the following: Major system loads.

Question #5 Given the following plant conditions:

• The unit is in MODE 3 preparing for startup

• The Startup Transformer, XMR01, is inadvertently deenergized

• NE02 Diesel fails to start Which of the following describes the pressurizer heater groups that are available for RCS pressure control?

A. Backup Heater Group "A" ONLY B. Backup Heater Group "B" ONLY C. Backup Heater Group "A" AND Variable Heater Group "C" D. Backup Heater Group "B" AND Variable Heater Group "C" Answer: A Explanation: The power supply lineup for mode 3 will be NB01 powered from the switchyard ring bus and NB02 powered from the Startup Transformer. Station Service Power (Pa busses) will be powered from the Startup Transformer also. Specifically, Backup Heater Group A is powered from NB01 via PG21 and Backup Heater Group B is powered from NB02 via PG22. The Variable Heater Group C is powered from PA02. For the given conditions above, PA02 losses power and since the B EDG, NE02, fails to start, NB02 losses power. Therefore, the only heater group available for RCS pressure control is Backup Heater Group A since NB01 will remain powered from the switchyard ring bus.

A. Correct.

B. Incorrect - NB02 becomes deenergized C. Incorrect - PA02 losses power causing a loss of PG24 losing the Variable Heater Group C D. Incorrect - NB02 becomes deenergized and PA02 losses power causing a loss of PG24 losing the Variable Heater Group C

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s):

(1) E-21001(Q), Main Single Line Diagram Rev 23 (2) Callaway FSAR Section 8, page 8.2-1 References to be provided to applicants during examination: None Learning Objective:

(1) T61.0110 LP-01, Switchyard - DRAW and EXPLAIN a one line diagram of the switchyard distribution system illustrating normal breaker alignment. Include:

• 345KV breakers

• Disconnects

• Buses A and B

• Safeguards transformers

• Ring bus

• Incoming and outgoing transmission lines (2) T61.0110 LP-06, Safeguards Power I. LIST the loads supplied by NB01 and NB02.

(3) T61.0110 LP-09, Reactor Coolant System

3. DESCRIBE the purpose and operation of the following RCS components to include interlocks, controller operations and power supply:

8. Pressurizer Heaters Question Source: Bank # ___X L15768______

Modified Bank # ______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 063 DC Electrical Distribution Group # 1 K/A # A3.01 Importance Rating 2.7 Ability to monitor automatic operation of the DC electrical system, including: Meters, annunciators, dials, recorders, and indicating lights.

Question #6 Given the following plant conditions:

• The plant is operating at 100% power when the following indications are observed by the crew:

Based on these conditions, 120V AC Bus NN01 is being supplied power from

____(1)____, and the crew will implement ______(2)_____ ?

A. (1) NK01 (2) OTO-NK-00001, Failure of NK Battery Charger B. (1) NK01

NRC Site-Specific Written Examination Callaway Plant Reactor Operator (2) OTO-NK-00002, Loss of Vital 125VDC Bus C. (1) NG01A (2) OTO-NK-00001, Failure of NK Battery Charger D. (1) NG01A (2) OTO-NK-00002, Loss of Vital 125VDC Bus Answer: A Explanation:

A. Correct. Based on the indications of the annunciators NN11 inverter trouble/transfer and NK01 Trouble, it can be determined that a malfunction on the DC electrical system has occurred.

With the indications of NK11 showing a discharge and bus voltage, with no current flow from the charger NK21, it is indicative of the battery supplying power to NK01 which in turn is supplying power to NN01 due the normal lineup. The indications of NK12 and NK02 are shown for comparison to normal values. Entry conditions for OTO- NK-00001, Failure of NK Battery Charger are met and entry to this procedure is required.

B. Incorrect. Indications show that NK01 still has bus voltage and therefore has not lost power, so entry into OTO-NK-00002, Loss of Vital 125VDC Bus is an incorrect action.

C. Incorrect. Plausible because NG01A is the alternate power supply through both the static transfer switch and the SOLA Transformer. In the event NK01 is lost, then NG01A will be supplying NN01 via the static transfer switch. If the static transfer switch is lost then the SOLA transformer can be placed in service to directly supply NN01 via a manual breaker transfer on NN01. See explanation A &B.

D. Incorrect. Plausible because NG01A is the alternate power supply through both the static transfer switch and the SOLA Transformer. In the event NK01 is lost, then NG01A will be supplying NN01 via the static transfer switch. If the static transfer switch is lost then the SOLA transformer can be placed in service to directly supply NN01 via a manual breaker transfer on NN01. See explanation A &B.

Technical Reference(s): OTO-NK-00001, Failure of NK Battery Charger REV 13 References to be provided to applicants during examination: None Learning Objective: T61.003B LP-B-26, Obj. B. Describe symptoms or entry conditions fo OTO-NK-00001, Failure of NK Battery Charger.

Question Source: Bank # __L16455____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

Used Bank Question, but added pictures for interpretation vice original question with bulleted information in stem.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 064 Emergency Diesel Generator Group # 1 K/A # K6.07 Importance Rating 2.7 Knowledge of the effect of a loss or malfunction of the following will have on the ED/G system: Air receivers.

Question #7 Given the following plant conditions:

• The plant is at 100% power.

• A relief valve has failed open on 'C' Starting Air Receiver for 'B' Emergency Diesel Generator (EDG).

• The leakage exceeds the capacity of the starting air compressor.

If an EDG start signal is generated, the B EDG ___(1)___ start, because the C and D air receivers ____(2)____ cross connected.

(1) (2)

A. will are B. will are not C. will not are D. will not are not Answer: B Explanation: The starting air compressor system for the B EDG consists of 2 Starting air tanks TKJ02C and TKJ02D. These are refered to as the C and D starting air receivers.

Each Starting air tank has an inlet check valve, KJV711B and KJV712B, that prevents a flaw / depressurization in one air tank from affecting the other tank. The discharge of the starting air tanks are NOT cross connected, i.e. KJV760B is closed.

A. Incorrect - Plausible if the candidate assumes that the air pressure in the D receiver will compensate for the loss of the C air receiver and the receivers are cross connected. The EDG will start because the Air receivers ARE NOT cross connected therefore the fault in the C air receiver will not affect the D air receiver allowing the EDG to start on only one receiver.

B. Correct - the D Air receiver has sufficient starting air pressure required for one start attempt per the Tech Spec basis. Prior to leak on the C Air receiver, it is assumed that starting receiver pressure was 610- 640 psig (normal band) The Air receivers are not cross connected during normal operation to ensure redundancy in the starting capability of the EDG. The inlet lines have

NRC Site-Specific Written Examination Callaway Plant Reactor Operator check valves which allow a single air compressor to supply both receivers simultaneously while still ensuring independence and redundancy of the system.

C. Incorrect. Plausible if the candidate assumes that both air receivers are required to supply sufficient starting air to the EDG with the system operated cross connected. If they assume they are operated cross connected they could also assume the fault in the C receiver is also degrading the starting pressure in the D receiver.

D. Incorrect. Plausible if the candidate assumes that both air receivers are required to supply sufficient starting air to the EDG with the system operated without the receivers cross connected.

Technical Reference(s): Tech Spec Bases 3.8.3 Condition E, M-22KJ02 Rev 20 References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-03 KJ NE Standby Generation Objective C: Describe the purpose, major components and operation of the following Standby Diesel Generator support systems: Air Start System Question Source: Bank # ___X__L16192_

Modified Bank # ______

New _______

Question History: Last NRC Exam __2005__________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.7 )

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 064 Emergency Diesel Generator Group # 1 K/A # K3.02 Importance Rating 4.2 Knowledge of the effect that a loss or malfunction of the ED/G system will have on the following: ESFAS controlled or actuated systems.

Question #8 Given the following plant conditions:

• The Plant was operating at 100% power when a LOCA occurred in conjunction with a Loss of Off-site Power

• E-1, Loss of Reactor or Secondary Coolant, is in progress

• B Emergency Diesel Generator NE02 TRIPS

• Containment pressure peaked at 28 psig

• ANN 47B, RWST LEV LOLO 2, is in alarm Which of the following describes the position of EN HV-1 (Train A) and EN HV-7 (Train B), Containment Recirc Sump to Containment Spray Pumps, at the completion of ES-1.3, Transfer To Cold Leg Recirculation?

EN HV-1 EN HV-7 A. OPEN OPEN B. OPEN CLOSED C. CLOSED OPEN D. CLOSED CLOSED Answer: B Explanation:

A. Incorrect. Plausible because under normal conditions with power available to both A and B train components, this would be correct. With the loss of NB02 due to the trip of the EDG, the B Train valves (including EN HV-7) and Spray Pump will not have power.

B. Correct. ES-1.3 will direct the operator to open both the containment sump recirculation valves. With the loss of NB02 due to the trip of the EDG, the B Train valves (including EN HV-7) and Spray Pump will not have power.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Incorrect. Plausible if the candidate incorrectly crosses the trains believing that the A train components are affected by the loss of the EDG. Incorrect because EN HV-1 (A train) will be aligned correctly to the Open position, and EN HV-7 (B Train) will not be realigned.

D. Incorrect. Plausible if the candidate determines that containment pressure is below the setpoint for containment spray actuation and therefore the spray pumps are not running and swap over for containment spray is not required. Incorrect because Containment spray actuates at 27 psig in containment and swapover to containment sump is required at the RWST Lo Lo 2 alarm setpoint.

Technical Reference(s): ES-1.3, Transfer to Cold Leg Recirculation, REV 11, Drawing E-23EN02 REV 13.

References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-18, Obj. C. Explain the interlocks, controls and power supplies to: 1) Containment Spray Pumps, 2) Containment Recirculation Sump Encapsulated Suction Valves, 3) Containment Spray Pump Discharge Valves.

Question Source: Bank # ______

Modified Bank # ___L15657___

New _______

Question History: Last NRC Exam ______N/A______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

K/A Match: Candidate must know that the Loss of the Emergency Diesel Generator results in a loss of the Vital Power Bus, and how the of the loss of the power to the bus will impact the ESFAS actuated Containment Spray System.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000038 Steam Gen. Tube Rupture / 3 Group # 1 K/A # EA1.19 Importance Rating 3.4 Ability to operate and monitor the following as they apply to a SGTR: MFW System status indicator.

Question # 9 Given the following plant conditions:

• The plant is operating at 100% power.

• The Crew is performing OTO-BB-00001, Steam Generator Tube Leak.

• The Balance of Plant is attempting to identify the affected Steam Generator.

Based on the above indications, which Steam Generator has experienced a tube leak?

A. A S/G B. B S/G C. C S/G D. D S/G Answer: B

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Explanation:

A. Incorrect. These are normal indications for feed flow, steam flow, and level at 100% power.

Plausible because of elevated level compared to D S/G with small Steam Flow Feedflow mismatch.

B. Correct. Based on Reduced SG Feedwater flow with a Stable Steam Generator Level, B S/G indicates that additional feedwater is coming from another source,(ie. the ruptured tube from the RCS)

C. Incorrect. Even Though C S/G indicates a higher level than the other generators, it is because feedflow is much higher than steam flow, so a higher level is expected.

D. Incorrect. If it is incorrectly assumed that level will lower due to a tube leak, then D S/G is plausible, however feedflow and steam flow are matched with level stable, indicating that level is simply lower than expected but responding normally.

Technical Reference(s): OTO-BB-00001, Steam Generator Tube Leak, REV 24 References to be provided to applicants during examination: None Learning Objective: T-61.003B, LP-B-10, Obj. B. Describe the symptoms or entry conditions for OTO-BB-00001, Steam Generator Tube Leak.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR 41.7)

Comments:

K/A Match: The candidate must interpret the Main Feedwater flow to each steam generator in order to see the lowered Feedwater flow to the steam generator with the tube rupture.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000055 Station Blackout / 6 Group # 1 K/A # EK1.01 Importance Rating 3.3 Knowledge of the operational implications of the following concepts as they apply to the Station Blackout : Effect of battery discharge rates on capacity.

Question #10 Given the following plant conditions:

• A Loss of All AC has occurred and the crew is performing ECA-0.0, Loss of All AC Power.

• NK12 battery discharge is indicating 170 amps.

Which of the following is the MAXIMUM time that NK02, Vital 125VDC Bus, is assumed to be able to supply power to the loads if the battery was fully charged initially?

A. 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> B. 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> C. 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> D. 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> Answer: A Explanation:

A. Correct. NK12 is rated for 900 amp hours and at a discharge rate of 170 amps the battery capacity is 5.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Therefore 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> is the maximum listed time the battery would be operable.

B. Incorrect. NK12 is rated for 900 amp hours, however PJ01 is rated at 1200 amp hours, and at a discharge rate of 170 amps the battery capacity is 7.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, Therefore 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> is plausible if the incorrect battery rating was used.

C. Incorrect. NK12 is rated for 900 amp hours, however NK11 is rated for 1650 amp hours, and at this discharge rate would be 9.7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />. Therefore 9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> is plausible if the incorrect battery rating was used.

D. Incorrect. NK12 is rated for 900 amp hours; however, PK12 is rated for 2400 amp hours. With a discharge rate of 170 amps per hour, this would equate to 14.1 hours1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. Therefore 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> is plausible if the incorrect battery rating was used.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s):

1. OTN-NK-00001, 125 VDC Bus NK02 and Distribution System, REV 2

2. E-21PJ01, Rev 4

3. E-21PK01, Rev 15

4. E-21NK02, Rev 8 References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-06, Obj. M. Explain the precautions and limitations and bases for the following components / conditions associated with OTN-NK-00001, Class 1E 125 VDC Electrical System: 1. Battery Capacity, 2. Maximum NK Battery Charge amperage output.

Question Source: Bank # ______

Modified Bank # ___L7911___

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR 41.8)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000056 Loss of Off-site Power / 6 Group # 1 K/A # AK3.01 Importance Rating 3.5 AK3.01 Knowledge of the reasons for the following responses as they apply to the Loss of Offsite Power: Order and time to initiation of power for the load sequencer Question #11 Given the following conditions:

• The plant tripped from 100% power due to loss of off-site power.

When will the A MDAFW pump start during this event and what is the reason for the starting sequence?

A. The A MDAFW pump will start after the So that at no time during the loading A ESW pump has started sequence does the voltage decrease to less than 75 percent of 4.16 kV.

B. The A MDAFW pump will start after the So that at no time during the loading A ESW pump has started sequence does the frequency decrease to less than 90 percent of 60 Hz.

C. The A MDAFW pump will start prior to So that at no time during the loading the A ESW pump if a MDAFAS is sequence does the voltage decrease to present. less than 75 percent of 4.16 kV.

D. The A MDAFW pump will start prior to So that at no time during the loading the A ESW pump if a MDAFAS is sequence does the frequency decrease present. to less than 90 percent of 60 Hz.

Answer: A Explanation:

Starting sequence: The Shutdown sequencer blocks all other AFW pumps starts until the MDAFW pumps are started at 30 seconds by the Shutdown sequencer. The A ESW pump is started at 20 seconds by the Shutdown Sequencer.

Reason: The diesel generators are designed as follows:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator

a. To start and accelerate to rated speed, in the sequence shown in Figure 8.3-2, all the needed engineered safety features and emergency hot shutdown loads.

b. So that at no time during the loading sequence do the frequency and voltage decrease to less than 95 percent of 60 Hz and 75 percent of 4.16 kV, respectively.

A. Correct. Per the explanation above this is the correct sequence, and reason.

B. Incorrect. Per the explanation above this is the correct sequence, however Frequency is allowed to drop during the starting of large loads to 95% of 60 Hz.

C. Incorrect. Per the explanation above this is the incorrect sequence. The Shutdown sequencer blocks ALL MDAFW pump starts until after the A ESW pump is started. The reason is correct D. Incorrect. Per the explanation above this is the incorrect sequence. The Shutdown sequencer blocks ALL MDAFW pump starts until after the A ESW pump is started. Also the reason is incorrect. Frequency is allowed to drop during the starting of large loads to 95% of 60 Hz.

Technical Reference(s): FASAR Section 8.1.4.3, 8.3.1.1.3, and Figure 8.3-2 References to be provided to applicants during examination: None Learning Objective: DISCUSS the reason for the five (5) second sequencing interval and LIST the major loads sequenced by the following:

1. LOCA Sequencer.

2. Shutdown Sequencer.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

CFR 45.7 Comments:

8.1.4.3 Design Criteria, Regulatory Guides, and IEEE Standards The diesel generators are designed as follows:

c. To start and accelerate to rated speed, in the sequence shown in Figure 8.3-2, all the needed engineered safety features and emergency hot shutdown loads.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator

d. So that at no time during the loading sequence do the frequency and voltage decrease to less than 95 percent of 60 Hz and 75 percent of 4.16 kV, respectively.

e. To recover from transients caused by step-load increases or resulting from the disconnection of full load so that the speed does not cause damage to moving parts.

During recovery, the speed of the diesel generator will not exceed 75 percent of the difference between nominal speed and the overspeed trip set point, or 115 percent of nominal, whichever is lower. Voltage will be restored to within 10 percent of nominal and frequency within 2 percent of nominal in less than 60 percent of each load sequence time interval.

8.3.1.1.3 Standby Power Supply Following diesel start and connection to the Class 1E bus, the loads are automatically sequenced onto the bus at programmed time intervals. A fast responding exciter and voltage regulator ensure voltage recovery of the diesel generator after each load step. Field flashing is utilized on the diesel generators for fast voltage buildup during the start sequence. Momentary voltage and frequency dips will not exceed a maximum of 25 percent below nominal rating (4.16 kV) for voltage and 5 percent for frequency.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Group # 1 000057 Loss of Vital AC Inst. Bus / 6 K/A # AA2.03 Importance Rating 3.7 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: RPS panel alarm annunciators and trip indicators.

Question #12 Given the following plant conditions:

• The plant is operating at 100% power

• A Loss of NN04, Vital AC Instrument Bus has occurred Which of the following Reactor Protective Systems bistable lights will be LIT on SB0069, Rx Trip/Bypass Permissive Status Panel?

A. Low RCS Flow B. Source Range Hi Flux C. Pressurizer Lo Pressure D. Pressurizer Hi Water Level Answer: C Explanation:

A. Incorrect. The three channels of RCS Lo Flow detector bistables are powered from NN01, NN02, and NN03. A loss of NN04 does not affect the RCS Low flow Reactor trip signal.

B. Incorrect. There are 2 channels of SR High Flux that input to the RPS system for a reactor trip. It could be incorrectly assumed that a loss of NN04 will cause a loss of power to the Channel 2 SRNI. This is plausible because many other systems utilize a power scheme of Train A off of channel 1, and Train B off of channel 4.

C. Correct. There are 4 channels for pressurizer pressure, therefore when NN04 is lost, the channel 4 of pressurizer pressure loses power and the bistable trips actuating channel 458C of the pressurizer pressure system.

D. Incorrect. The three channels of Pressurizer Hi Water Level Reactor Trip signals are powered from NN01, NN02, and NN03. A loss of NN04 does not affect the Pressurizer Hi Water Level Reactor Trip Signal.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): OTO-NN-00001, Loss of Safety Related Instrument Power, REV 32 References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-27, Obj. K. Identify the RPS Main Control board controls, alarms and indications and Explain how each is used to predict, monitor, or control the RPS.

Question Source: Bank # ______

Modified Bank # ______

New ____X___

Question History: Last NRC Exam ___N/A_________

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 43.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000058 Loss of DC Power / 6 Group # 1 K/A # 2.2.37 Importance Rating 3.6 Ability to determine operability and/or availability of safety related equipment.

Question #13 Given the following plant conditions:

• The plant is operating at 100% power, when the following annunciator is received:

o 28C, NK04 TROUBLE

• NK04 voltage indication is 0 volts.

Which of the following describes the long term effect on the plant?

A. The Reactor will trip, EDG "B" is NOT available for use.

B. The Reactor will remain at power, EDG "B" is NOT available for use.

C. The Reactor will remain at power, Turbine Driven AFW Pump will automatically start.

D. The Reactor will trip, the Turbine Driven AFW Pump must be Manually started if required.

Answer: A Explanation:

A. Correct. Due to MISVs and FWIV failing closed on a loss of NK04, the reactor will be tripped.

B EDG was not initially running, therefore it is NOT available due to a loss of field flash capability.

B. Incorrect. Due to MISVs and FWIV failing closed on a loss of NK04, the reactor will be tripped. B EDG was not initially running, therefore it is NOT available due to a loss of field flash capability.

C. Incorrect. Due to MISVs and FWIV failing closed on a loss of NK04, the reactor will be tripped. If NK02 is lost, the TDAFW pump governor control valve DC power is lost and may need to be started manually if it is required.

D. Incorrect. Due to MISVs and FWIV failing closed on a loss of NK04, the reactor will be tripped. If NK02 is lost, the TDAFW pump governor control valve DC power is lost and may need to be started manually if it is required.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): OTO-NK-00002, Loss of Vital 125VDC Bus, REV 13 References to be provided to applicants during examination: None Learning Objective: T61.003B LP-B-50, Obj. D. Discuss the major cautions and notes contained in the body and attachments of OTO-NK-00002, Loss of Vital 125 VDC Bus.

Question Source: Bank # ______

Modified Bank # ___L16650___

New _______

Question History: Last NRC Exam ______Callaway 2011 ILT Audit Exam______

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000062 Loss of Nuclear Svc Water / 4 Group # 1 K/A # AA2.02 Importance Rating 2.9 Ability to determine and interpret the following as they apply to the Loss of Nuclear Service Water: The cause of possible SWS loss.

Question #14 Given the following plant conditions:

• The plant was operating at 100% power when a LOCA occurred.

• The Balance of Plant is performing E-0, Reactor Trip or Safety Injection, Attachment A and observes the following indications:

• Annunciator 55A, ESW B PRESS LO - FLOW LO is LIT Which of the following is the cause of the Annunciator 55A?

A. B ESW Pump shaft shear B. B ESW Pump failed to start C. B ESW Pump Discharge Strainer is clogged.

D. B Train Service Water Cross Connect Valve failed CLOSED

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Answer: A Explanation:

A. Correct. Based on the indications of a Low Flow and Low Pressure on the B Train ESW system, and a Red Running light on the B EWS pump, this indicates that the pump has started and is running but not producing any discharge flow or pressure. This is characteristic of a sheared shaft on the ESW pump.

B. Incorrect. Upon initiation of the Safety Injection the ESW pumps will Auto Start on the Timed Sequencer. The RED indicating light indicates that the breaker for the B ESW pump has started.

The annunciator 55A which is driven off of the ESW pump discharge pressure will only activate if the pump is running and pressure is low.

C. Incorrect. Plausible because if the pump is running and the strainer is clogged, then low flow will occur on the system, however the pressure detector is located on the discharge of the pump, upstream of the strainer, and will indicate a higher than normal pressure. Annunciator 55C ESW Strainer D/P HI would also be in alarm, not provided in the stem.

D. Incorrect. The Service Water Cross Connect valves are normally OPEN supplying the ESW loads while the ESW pumps are in Standby. On initiation of an ESW pump start, these valves will automatically CLOSE to isolate the Service Water System from the ESW system. The SI signal can be validated with this picture using both the Service Water Cross Connect valves being closed and the UHS Return valves being OPEN. If one of these valves closed under normal operation, It will cause Annunciator 55A to come in, however the ESW pump will not Automatically start.

Technical Reference(s): Drawing M-U2EF01, Piping & Instrumentation Diagram Essential Service Water System REV 65, Drawing M-22EF01, Piping & Instrumentation Diagram Essential Service Water System REV 78 References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-05, Obj. E. IDENTIFY the Essential Service Water System Main Control Board (MCB) controls, alarms and indications and DESCRIBE how each is used to predict, monitor or control changes in the Essential Service Water System.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam __N/A__________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 43.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 W/E04 LOCA Outside Containment / 3 Group # 1 K/A # EK1.2 Importance Rating 3.5 Knowledge of the operational implications of the following concepts as they apply to the (LOCA Outside Containment): Normal, abnormal and emergency operating procedures associated with (LOCA Outside Containment).

Question #15 Given the following plant conditions:

• A LOCA outside containment has occurred.

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

(1) Which of the following actions will be the FIRST to be attempted to isolate the break; And (2) Which indication is used to determine if the leak has been isolated in accordance with ECA-1.2?

A. (1) Isolate RHR piping (2) RCS pressure B. (1) Isolate RHR piping (2) Pressurizer level C. (1) Isolate the Boron Injection Header (2) RCS pressure D. (1) Isolate the Boron Injection Header (2) Pressurizer level Answer: A Explanation:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator A. Correct. ECA-1.2 directs isolating each system that is connected to the RCS and is located outside of containment, starting with the RHR system because it is a low pressure system connected to the high pressure RCS system and is therefore the most probable location for a LOCA outside containment. RCS Pressure is monitored because pressure will rise when the break is isolated.

B. Incorrect. Isolating RHR piping first is correct, however Pressurizer level is not monitored while in ECA-1.2. Under normal conditions, RCS pressure rising will provide faster response to the leak isolation. This is plausible because Pressurizer level is monitored for other RCS leak procedures to determine SI criteria.

C. Incorrect. ECA-1.2 directs closing and opening normally open valves in systems that are connected to the RCS. The Boron Injection Header is isolated last because it is the highest pressure rated system and therefore the least likely to develop a leak. See A explanation.

D. Incorrect. See B & C.

Technical Reference(s): ECA-1.2, LOCA Outside Containment, Rev 7 References to be provided to applicants during examination: None.

Learning Objective: T61.003D LP-D-14, Obj. E. State and Explain the parameters which are evaluated, including their Criteria and Basis, to transition from ECA-1.2 to other procedures. F.

Outline the procedural flowpath including major systems and equipment operations in accomplishing the goal of ECA-1.2.

Question Source: Bank # __ L16515____

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2011 Audit Exam_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.8)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 W/E11 Loss of Emergency Coolant Recirc. / 4 Group # 1 K/A # EK2.1 Importance Rating 3.6 Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Question #16 Given the following plant conditions:

• The plant was operating at 100% power when a LOCA occurred.

• Due to equipment failures, the crew is performing actions contained in ECA-1.1, Loss of Emergency Coolant Recirculation.

• The RO momentarily presses the SI (RWST) Switchover Signal RESET pushbuttons.

Which of the following describes the function provided by pressing the SI (RWST) RESET pushbuttons?

A. Restores automatic control of the RHR Pump containment recirc valves.

B. Prevents inadvertent loss of RWST inventory due to automatic switchover.

C. Allows Containment Spray Pump discharge valves to be closed when no longer required.

D. Allows SI and RHR pumps to be started and stopped as required during the Cold Leg Recirculation phase of the event.

Answer: B Explanation:

A. Incorrect. This function is Removed by resetting the RWST SI signal. With an SI present when the RWST reaches 36% RHR suction valves will automatically swap over to the containment sump allowing the maximum volume of the RWST to be transferred to the sump.

B. Correct. The purpose of resetting the signal is to prevent inadvertent loss of RWST inventory due to an automatic swapover. This prevents loss of RWST inventory to the containment sump and possible ECCS pump damage.

C. Incorrect. This function is allowed when the Containment Spray Actuation Signal (CSAS) signal is reset, but is not activated on an SI nor is reset by the SI reset switches or the RWST SI reset switches.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator D. Incorrect. Resetting the overall SI Signal with the SI Reset Switches (Different than the RWST SI switches) allows the starting and stopping of ECCS pumps during subsequent recovery phases of a LOCA.

Technical Reference(s): ECA-1.1, Loss of Emergency Coolant Recirculation REV 10, Background Document for ECA-1.1, Loss of Emergency Coolant Recirculation REV 5.

References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-13, Obj. C. Identify RWST Main Control Board controls, alarms, and indications and Describe how each is used to predict, monitor or control changes in the RWST.

Question Source: Bank # ___L16244___

Modified Bank # ______

New _______

Question History: Last NRC Exam ______Callaway 2007 ILT NRC Exam______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 BW/E04; W/E05 Inadequate Heat Transfer - Loss Group # 1 of Secondary Heat Sink / 4 K/A # 2.4.21 Importance Rating 4.0 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Question #17 Given the following plant conditions:

• The plant was operating at 100% power when a Steamline Low Pressure Safety Injection occurred which resulted in a reactor trip at 0734.

• The crew transitioned from E-0, Reactor Trip Or Safety Injection, to E-2, Faulted Steam Generator Isolation, and is isolating the faulted Steam Generator.

The following plant conditions exist at 0753:

• RCS Cold Leg Temperatures A - 423°F, B - 420°F C - 431°F, D - 232°F

• Containment Pressure 32 psig

• Containment Spray Pumps None Running

• RCS WR Pressure 1125 psig

• PZR Level Off Scale Low

• Source Ranges 100 cps - stable

• All S/G Levels (NR) 0%

• AFW Flow 280,000 lbm/hr Given the above conditions, what is the HIGHEST Safety Function that is challenged?

A. Subcriticality B. Heat Sink C. Integrity D. Containment

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Answer: B Explanation:

A. Incorrect. Reactor power indicates less than 5% therefore FR-S.1 entry conditions are not met. Therefore Subcriticality is not challenged B. Correct. With Narrow Range S/G levels less than 7% and total Aux Feedwater Flow less than 285,000 lbm/hr, conditions are met for a Red Path Loss of Secondary Heatsink and FR-H.1 entry conditions are met. The crew has transitioned out of E-0 and status tree monitoring has begun and the transition can be made, prior to this no transition can be made unless directed by E-0.

Heat sink is therefore the highest safety function that is challenged C. Incorrect. Conditions are met for an Orange Path on FR-P.1 due to the cold leg temperature on D loop, however the higher path of Red and the higher status of FR-H.1 make the FR-H.1 transition a higher priority and the correct path. However, the heat sink safety function is higher D. Incorrect. Conditions are met for an Orange Path on FR-Z.1 due to the high containment pressure and no spray pumps running, however the higher path of Red and the higher status of FR-H.1 make the FR-H.1 transition a higher priority and the correct path. However, the heat sink safety function is higher Technical Reference(s): CSF-1, Critical Safety Function Status Trees, REV 10 References to be provided to applicants during examination: None Learning Objective: T61.003D LP-D-26, Obj. B. Describe the Symptoms and/or Entry conditions for: 1. FR-H.1, Response to Loss of Secondary Heat Sink.

Question Source: Bank # ___

Modified Bank # R8481_________

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.10)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000077 Generator Voltage and Electric Group # 1 Grid Disturbances / 6 K/A # 2.1.27 Importance Rating 3.9 Knowledge of system purpose and/or function.

Question #18 What components normally protect the Engineered Safety Feature buses from fluctuating switchyard voltages?

A. Manual load tap changers ONLY B. Automatic load tap changers ONLY C. Manual load tap changers AND associated capacitor banks D. Automatic load tap changers AND associated capacitor banks Answer: D Explanation: Per the FSAR chapter 8 and TS bases 3.8.1, Automatic load tap changers associated with the ESF transformers, as well as associated capacitor banks, provide voltage regulation for the preferred sources in the event of changing switchyard voltage.

A. Incorrect. The manual tap changers are a backup to the automatic tap changers if the auto tap changers fail.

B. Incorrect. The ESF buses are protected by BOTH the auto tap changers and the capacitor banks.

C. Incorrect. The manual tap changers are a backup to the automatic tap changers if the auto tap changers fail.

D. Correct. The ESF buses are protected by BOTH the auto tap changers and the capacitor banks.

Technical Reference(s): TS Bases 3.8.1 and FSAR chapter 8 References to be provided to applicants during examination: None Learning Objective: T61.0110, LP-01, Obj. D. IDENTIFY locations of controls and indications on the main control board and describe how each may be used to predict monitor, or control changes in the switchyard distribution system.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam __N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

Offsite power is supplied to the unit switchyard from the transmission network by four transmission lines. From the switchyard, two electrically and physically separated circuits provide AC power, through ESF transformers, to the 4.16 kV ESF buses. Automatic load tap changers associated with the ESF transformers, as well as associated capacitor banks, provide voltage regulation for the preferred sources in the event of changing switchyard voltage. A detailed description of the offsite power network and the circuits to the Class 1E ESF buses is found in the FSAR, Chapter 8 (Ref. 2).

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000024 Emergency Boration / 1 Group # 2 K/A # AK2.01 Importance Rating 2.7 Knowledge of the interrelations between Emergency Boration and the following: Valves Question #19 Given the following plant conditions:

• The plant is at 100% power

• A transient occurs requiring Emergency Boration to be initiated in accordance with OTO-ZZ-00003, Loss of Shutdown Margin

• Emergency Boration flow is 20 gpm due to a clogged boric acid filter

• Attempts to bypass the filter have been unsuccessful.

Which of the following valves must be opened in order to successfully provide the required Emergency Boration flow?

A. BG FCV-110B, Makeup To VCT Outlet B. BN HV-8806A, RWST to SI Pumps suction valve C. BN LCV-112D, Charging Pump Suction From RWST D. BG HV-8104, Emergency Borate To Charging Pump Suction Answer: C A. Incorrect. Plausible because it is the second attempted path to supply boric acid to the charging pump suction after the emergency borate valve is opened. Incorrect because this is downstream of the boric acid filter ,and will not provide adequate flow if the filter is clogged.

B. Incorrect. Plausible because this is the RNO action from step 4 in check the charging system available. Nothing tells the candidate that the charging system is unavailable. the RNO says to perform Attachment B. This is the first valve that is opened in Attachment B.

C. Correct. If attempts to supply boric acid from the BAST have been unsuccessful as indicated by Emergency Boration flow being less than 30 gpm, then the emergency boration should be accomplished from the RWST by opening BN LCV-112D, CCP suction from the RWST.

D. Incorrect. Plausible because it is the first attempted path to supply boric acid to the charging pump suction after it is determined that an emergency boration is required. Incorrect because this is downstream of the boric acid filter ,and will not provide adequate flow if the filter is clogged.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): OTO-ZZ-00003, Loss of Shutdown Margin, Rev 18 References to be provided to applicants during examination: None Learning Objective: C DRAW, LABEL and EXPLAIN a one line diagram of the CVCS to include the components listed in Objective B showing system alignment under any of the following conditions:

1. Normal Operation

2. Safety Injection

3. Containment Isolation Signal Phase A (CISA)

4. Cold Leg Recirculation Question Source: Bank # __L16571____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____2013 NRC________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000028 Pressurizer Level Malfunction / 2 Group # 2 K/A # 2.4.46 Importance Rating 4.2 Ability to verify that the alarms are consistent with the plant conditions.

Question #20 Given the following plant conditions:

• The plant is operating at 100% power.

• The upper selected Pressurizer level channel, BB-LI-459A, has failed to 0%.

• Annunciator 32C, PZR Lo Lev Dev, is LIT.

Based on these plant conditions, the Reactor Operator will observe which of the following?

PZR Variable Heaters Charging Flow A. ON Lowering B. ON Rising C. OFF Lowering D. OFF Rising Answer: D A. Incorrect. Plausible because if 459A is NOT the controlling channel and BB-LI-459A fails, it will not provide input into any control circuitry and will NOT give any annunciators. These indications are accurate if BB-LI-459A was to fail high, all pzr heaters will energize and charging flow will reduce.

B. Incorrect. Plausible because the operator could assume that the increased charging flow due to the failed instrument will drive pressurizer level up and require additional heat input from the heaters due to the insurge into the pressurizer. Incorrect because pressurizer heaters will turn off when annunciator 32C is LIT in order to prevent uncovering the pressurizer heaters, due to the assumption that Pressurizer level is low.

C. Incorrect. Plausible because when Annunciator 32C is LIT, then Letdown is isolated and pressurizer heaters will turn off, the candidate could assume that because it is an instrument failure, the backup channel will see actual level and reduce charging flow in order to prevent an undesired increase in actual pressurizer level following the isolation of letdown flow. Incorrect

NRC Site-Specific Written Examination Callaway Plant Reactor Operator because the same circuit that is controlling letdown and pressurizer heaters will also control charging flow.

D. Correct. Annunciator 32C is driven off of the channel selected for control of the pressurizer level. Therefore, with BB-LI-459A indicating 0% and Annunciator 32C LIT, it can be assumed that it was the upper controlling channel and upon failing low, will isolate letdown and turn off the PZR heaters which will drive charging flow to rise to restore pressurizer level. BB-LI-459A can only be selected as the upper controlling channel.

Technical Reference(s): OTO-BG-00001, Pressurizer Level Control Malfunction, Rev 20 References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-09, Obj. H. Identify the RCS Main Control Board controls, alarms, and indications, and Describe how each is used to predict, monitor, or control changes in the RCS.

Question Source: Bank # __16297____

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis _X____

10 CFR Part 55 Content:

(CFR: 41.10)

Comments:

K/A Match: Candidate must determine the expected conditions that are associated with receipt of a control room annunciator.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000032 Loss of Source Range NI / 7 Group # 2 K/A # AA2.02 Importance Rating 3.6 Ability to determine and interpret the following as they apply to the Loss of Source Range Nuclear Instrumentation: Expected change in source range count rate when rods are moved.

Question #21 The Unit is in Hot Shutdown making preparations for a reactor startup:

• Tavg is 557 °F.

• RCS boron is 1120 ppm.

• Initial Source Range counts are:

o N-31 counts are 90 cps o N-32 counts are 85 cps

• All Rods are inserted.

• Shutdown 7 days ago The shutdown banks have been fully withdrawn, and the following Source Range data is recorded:

• N-31 counts are 210 cps

• N-32 counts are 350 cps Based on the above indications, which of the following BEST describe the Source Range data that was collected and the potential failure?

A. N31 is indicating an abnormally Pulse height discriminator failed to low count rate a higher value B. N31 is indicating an abnormally Pulse height discriminator failed to low count rate a lower value C. N32 is indicating an abnormally Pulse height discriminator failed to high count rate a higher value D. N32 is indicating an abnormally Pulse height discriminator failed to high count rate a lower value Answer: D Explanation: Per OTG-ZZ-0001A source range channel count rate should not increase by a

NRC Site-Specific Written Examination Callaway Plant Reactor Operator factor of 4. The given counts rate in the stem N-31 was 90 cps. 90 cps x 4 = 360 cps. Based on this the operator should not expect counts to increase above 360 cps during the withdrawal of the Shutdown Bank. Likewise, N-32 was 85 cps. X4 = 340 cps. The operator should not expect counts to increase above 340 cps during the withdrawal of Shutdown banks.

If the pulse height discriminator fails high it would NOT count all of the pulses and would give a lower than expected indication.

If the pulse height discriminator fails low it would count more pulses and would give a higher than expected reading A. Incorrect - N31 is indicating less than 4 times the original value so this is a normal value. If the pulse height discriminator fails high it would NOT count all of the pulses and would give a lower than expected indication.

B. Incorrect - N31 is indicating less than 4 times the original value so this is a normal value. If the pulse height discriminator fails low it would count more pulses and would give a higher than expected reading C. Incorrect - N32 is indicating more than 4 times the initial value so this is an abnormal high count rate. If the pulse height discriminator fails high it would NOT count all of the pulses and would give a lower than expected indication.

D. Correct - N32 is indicating more than 4 times the initial value so this is an abnormal high count rate. If the pulse height discriminator fails low it would count more pulses and would give a higher than expected reading Technical Reference(s): OTG-ZZ-0001A References to be provided to applicants during examination: None Learning Objective: B. APPLY the requirements of the Precautions and Limitations of OTG-ZZ-0001A, to include:

1. Requirements for Control Rods, Shutdown Rods and Boron Concentration and when these requirements apply.

Question Source: Bank # ______

Modified Bank # ___X___

New _______

Question History: Last NRC Exam ____________, Modified from the 2012 NRC Surry RO written exam, http://pbadupws.nrc.gov/docs/ML1226/ML12269A249.pdf question 25.

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator (CFR: 43.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000037 Steam Generator Tube Leak / 3 Group # 2 K/A # AA1.04 Importance Rating 3.6 Ability to operate and / or monitor the following as they apply to the Steam Generator Tube Leak: Condensate air ejector exhaust radiation monitor and failure indicator .

Question #22 Given the following plant conditions:

• The plant is at 100% power

• Containment Mini-Purge is in Service

-13

• GT-RE-27, Fuel/Aux Building Exhaust Monitor indicates 3.35 E µCi/ml and STEADY

• GE-RE-92, Condenser Air Discharge Monitor indicates 55 µCi/ml and RISING

• RCS Tavg is 585F and STEADY

• Pressurizer level is 56% and stable

• Charging Flow is slowly rising

• RCS Iodine-131 last TWO samples each indicate 23 µCi/ml Which of the following describes the event in progress?

A. Fuel Element Failure B. Steam Generator Tube Leak C. RCS Leak outside of Containment D. Pressurizer Level Control Malfunction Answer: B Explanation:

A. Incorrect. Two consistent samples of RCS indicate constant Iodine-131 level of 23 µCi/ml.

This is not elevated enough nor is there a spike in activity which would be indicative of a fuel element failure.

B. Correct. The elevated value on the Condenser Air Discharge Rad Monitor GE-RE-92 is an indication of a Steam Generator Tube leak. .

C. Incorrect. GT-RE-27 indicates a steady reading at a normal level. If an RCS leak existed outside of containment this detector would show an increase in radiation levels. This is a

NRC Site-Specific Written Examination Callaway Plant Reactor Operator plausible distractor based on the student must know that GE-RE-92 indicates a SG tube leak instead of an RCS leak outside of Containment..

D. Incorrect. With Tavg constant and Pressurizer level stable, and charging rising, it is plausible that a malfunction is occurring with the level control system, however level is being maintained a charging rises due to the lost inventory through the Steam Generator Tube leak.

Technical Reference(s): OTO-BB-00001, Steam Generator Tube Leak, Rev 25 References to be provided to applicants during examination: None Learning Objective: T61.003B LP-B-10, Obj. B. Describe the symptoms or entry conditions for OTO-BB-00001, Steam Generator Tube Leak.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ______N/A______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000051 Loss of Condenser Vacuum / 4 Group # 2 K/A # 2.1.20 Importance Rating 4.6 Ability to interpret and execute procedure steps.

Question #23 Given the following plant conditions:

• Reactor power is at 70% and constant.

• Condenser backpressure has been steadily rising and is now 5.4" HgA.

• The crew has entered OTO-AD-00001, Loss of Condenser Vacuum, and just commenced a load reduction at 1% per minute.

15 minutes later, the following plant conditions exist:

• Condenser backpressure is now 7.8" HgA and constant.

Which of the following actions should be taken IAW OTO-AD-00001?

A. Trip the Reactor. Transition to E-0, Reactor Trip or SI.

B. Verify Control Rods are inserting to control Tavg to Tref to within 1.5°F.

C. Trip the Turbine. Transition to OTO-AC-00001, Turbine Trip Below P-9.

D. Continue to reduce Turbine load at less than or equal to 5% per minute.

Answer: A Explanation: With the initial conditions given, the crew is performing step 6 reducing load.

15 minutes later the plant condition have degraded requiring the RNO for continuous action step #1 to be implemented. With reactor power now at 55%, the correct action is to perform the RNO with Reactor Power greater than or equal to 10% and trip the reactor and transition to E-0.

A. Correct - See explanation above B. Incorrect - Step #12 which is a continuous action step requires to verify within 3°F not 1.5°F.

Furthermore, performing this action would mean that the candidate has not recognized the requirement to trip the reactor.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Incorrect - If the candidate does not correctly implement the continuous action step #1 (

candidate does not apply the 10% power criteria) they may believe incorrectly that a transition to OTO-AC-00001 is correct. Automatic turbine trip on condenser backpressure does not occur until 8.5 HgA.

D. Incorrect - A reactor trip and transition to E-0 are required. A reactor trip will trip the turbine.

Technical Reference(s): OTO-AD-00001, Loss of Condenser Vacuum, Revision 30 References to be provided to applicants during examination: None Learning Objective:

(1) T61.003B Off Normal Operations, LP B-07 Objective D. Given a set of plant conditions or parameters indicating a Loss of Condenser vacuum, ANALYZE the correct procedure(s) to be utilized and the required actions to stabilize the plant.

(2) T61.003B Off Normal Operations, LP B-07 Objective C DESCRIBE Continuous Action Step(s) including the required Response Not Obtained actions.

Question Source: Bank # ______

Modified Bank # ______

New _X______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis ___ __

10 CFR Part 55 Content:

CFR: 41.10 / 45.12 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000060 Accidental Gaseous Radwaste Group # 2 Rel. / 9 K/A # 2.2.38 Importance Rating 3.6 Knowledge of conditions and limitations in the facility license.

Question #24 Which of the following is the MAXIMUM allowable WHOLE BODY dose from Noble Gas at the site boundary that should be received by a person following an inadvertent release from a Radioactive Gas Storage tank?

A. 0.5 rem B. 1.5 rem C. 3 rem D. 5 rem Answer: A Explanation:

A. Correct. Restricting the quantity of radioactivity contained in each gas storage tank provides assurance that in the event of an uncontrolled release of the tank's contents, the resulting whole body exposure to a MEMBER OF THE PUBLIC at the nearest SITE BOUNDARY will not exceed 0.5 rem.

B. Incorrect. Plausible because 1.5 rem/yr is the annual exposure limit for Iodine-131 & 133, and Tritium to any organ, but is not the limit for noble gases.

C. Incorrect. Plausible because 3 rem/yr is the annual exposure limit at the site boundary for exposure to the skin for Noble Gas, but is not the limit for Whole Body.

D. Incorrect. Plausible because 5 mrem is the limit for dose to a member of the public to any organ due to LIQUID effluent release to unrestricted areas.

Technical Reference(s): FSAR 16.11.2.8, Gas Storage Tanks Limiting Condition for Operation.

References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-16, Obj. I. Describe the 16 FSAR Radwaste System Technical Specifications.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.10)

Comments:

K/A Match: Candidate must have knowledge of the limit contained in the FSAR Chapter 16 for Gaseous Radwaste Decay Tanks to determine the maximum limit of dose for off site personnel in the event of an accidental release from the tank.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000069 (W/E14) Loss of CTMT Integrity Group # 2

/5 K/A # EA1.3 Importance Rating 3.3 Ability to operate and / or monitor the following as they apply to the (High Containment Pressure): Desired operating results during abnormal and emergency situations.

Question #25 Given the following plant conditions:

• The Plant was operating at 100% power when a LOCA occurred.

• The RO is performing E-0, Rx Trip or Safety Injection, Attachment A, Automatic Action Verification, Step A8, Check If Containment Spray Should be Actuated, and observes the following indications:

Which of the following actions is required to be taken in response to these indications?

A. Manually Actuate CSAS Signal on Both Trains.

B. Place B Containment Spray Pump in Pull-to-Lock.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Open B Containment Spray Pump Discharge Valve.

D. Transfer Containment Spray Suction to the Containment Sump.

Answer: C Explanation:

A. Incorrect. E-0 Attachment A checks if Containment Spray Pumps are running, and if NOT the RNO states to Manually Actuate CSAS. Indications show the spray pump is running with indications of greater than 27 psig in containment.

B. Incorrect. Containment Spray Pumps are not placed in Pull-To-Lock unless damage to the pump is imminent or it is a hazard to personnel. All indications provided show proper operation.

If it is desired to secure the pump, CSAS is RESET and the handswitch is placed in Normal After Stop position.

C. Correct. Indications of High Discharge Pressure with no flow, and the Green light on the Discharge Valve indicate it failed to open. E-0 Attachment A RNO for valves not properly aligned states to Align CSAS valves as necessary. The Discharge Valves should be opened from the control board.

D. Incorrect. Transfer to containment sump is not required as indicated by the RHR pump still aligned to the RWST, the Automatic Swap over at 36% in the RWST has not yet occurred therefore swapping the Containment Spray is not yet required. Required at 12%level in the RWST, and indications are not consistent with pump cavitation due to loss of suction.

Technical Reference(s): E-0, Reactor Trip or Safety Injection. REV 16 References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-18 Obj. C. Explain the interlocks, controls and power supplies to: 1. Containment Spray Pumps, 2. Containment Recirculation Sump Suction Valves, 3.

Containment Spray Pump Discharge Valves.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator (CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 003 Dropped Control Rod Group # 2 K/A # AK1.02 Importance Rating 3.1 Knowledge of the operational implications of the following concepts as they apply to Dropped Control Rod: Effects of turbine-reactor power mismatch on rod control Question #26 Given the following plant conditions:

• The plant is operating at 40% power.

• Tavg is 558°F and stable.

• Tref is 559° F and stable:

• All control systems are aligned in automatic.

• Control Bank D Control Rods are at 150 steps.

Which of the following describes the plant response after a dropped rod and the required operator action in accordance with OTO-SF-00001, Rod Control Malfunctions?

Plant Response__ _ _____Required Action____

A. The difference between Tavg and Rods are placed in Manual Tref is smaller B. The difference between Tavg and Rods are left in Auto Tref is smaller C. The difference between Tavg and Rods are placed in Manual Tref is larger D. The difference between Tavg and Rods are left in Auto Tref is larger Answer: C Explanation:

A. Incorrect - 1st part is incorrect. However, the applicant understands when Tavg lowers, then Pstm going to the Turbine lowers (Tref). The effect of the lowering Tavg will be greater than Tref.

Also plausible because the immediate action for a Dropped RCC is to place the rods in MAN.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Incorrect - 1st part is incorrect. However, the applicant understands when Tavg lowers, then Pstm going to the Turbine lowers (Tref). The effect of the lowering Tavg will be greater than Tref.

Also plausible because the rod control demand is to have rods step out, but due to auto rod withdrawl being disabled the rods do not move. Without rod movement, the applicant leaves rods in Auto which is incorrect.

C. Correct - On a dropped rod, Tavg will lower. Rods are placed in manual to stabilize the plant and ensure SDM is maintained D. Incorrect - On a dropped rod, Tavg will lower. Rods are placed in manual to stabilize the plant and ensure SDM is maintained Technical Reference(s): OTO-SF-00001, Rod Control Malfunctions, Rev 15 References to be provided to applicants during examination: None Learning Objective: Given a Rod Control Malfunction, DETERMINE the affect the malfunction will have on various control circuits or systems and DISCUSS the impact on continued at-power operation.

Question Source: Bank # ______

Modified Bank # ___X___

New _______

Question History: Last NRC Exam ____________ Modified 2011 Turkey Point NRC exam http://pbadupws.nrc.gov/docs/ML1216/ML12166A066.pdf question 19 Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR: 41.8)

Comments:

This question matches the K/A in that it tests the impact that a dropped rod has on the power mismatch circuit of the Rod Control System.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 CE/A11; W/E08 RCS Overcooling - PTS / 4 Group # 2 K/A # EK2.2 Importance Rating 3.6 Knowledge of the interrelations between the (Pressurized Thermal Shock) and the following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Question #27 Which of the following mitigates a Pressurized Thermal Shock condition?

A. From the MCB, close an open normal PZR spray valve.

B. Direct an operator to locally isolate a stuck open S/G ASD.

C. From the MCB, close the block valve for a stuck open PZR PORV.

D. Direct an operator to locally isolate a stuck open Feedwater Reg Valve.

Answer: B Explanation: The correct answer is the only actions that moves the plant away from the condition of PTS A. . Incorrect. Plausible because pressure is a concern in a PTS condition, however if the spray valve is open it would be reducing pressure which is the opposite of a PTS condition that is high pressure and low temperature.

B Correct. Pressurized thermal shock occurs when a low temperature and high pressure condition is achieved in the primary system. By closing a stuck open S/G ASD, the operator will be stopping an uncontrolled cooldown which could result in a PTS condition.

C. Incorrect. Plausible because pressure is a concern in a PTS condition and the candidate could incorrectly assume that lowering pressure is placing the plant closer to a PTS condition, however this is the opposite of the conditions for a PTS to occur with high pressure and low temperature.

D. Incorrect. Plausible if the candidate assumes that a stuck open FW reg valve will still allow feedwater to be admitted to the steam generators and thus causing a cooldown. Incorrect because the FWIVs are utilized to isolate a steam generator, and feedwater flow is isolated to the steam generator by the closing of the FWIVs, so isolating the failed FW reg valve will provide additional protection but not have a direct impact on the condition of the plant.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): FR-P.1, Response to Pressurized Thermal Shock, Rev 10.

References to be provided to applicants during examination: None Learning Objective: T61.003D LP #28 A EXPLAIN the Purpose and Major Action Categories of:

1. FR-P.1, Response To Imminent Pressurized Thermal Shock Condition.

2. FR-P.2, Response To Anticipated Pressurized Thermal Shock Condition.

Question Source: Bank # __L16334____

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.7)

Comments: Used on the 2007 Audit exam

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 003 Reactor Coolant Pump Group # 1 K/A # A1.09 Importance Rating 2.8 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RCPS controls including: Seal flow and D/P Question #28 Given the following plant conditions:

• The Plant is in Mode 3 making preparations for a refueling outage.

• All RCPs are in Service.

• Waste Gas has been aligned to lower VCT Pressure.

• The Reactor Operator has been directed to monitor RCP parameters during this evolution.

VCT Pressure should NOT be lowered less than 15 psig with RCPs in operation in order to prevent _____(1)_____, and as VCT pressure is lowered, #1 Seal Return flow is expected to ____(2)_____.

A. (1) RCP seal damage (2) rise B. (1) RCP seal damage (2) lower C. (1) gases from coming out of solution at the charging pump suction (2) rise D. (1) gases from coming out of solution at the charging pump suction (2) lower Answer: A Explanation:

A. Correct. If VCT pressure is NOT between 15 psig and 70 psig when Reactor Coolant Pumps are in service, RCP seal damage may occur. VCT pressure is used to provide back pressure to the RCP #1 Seal Return, As VCT pressure is lowered, the D/P between the #1 Seal Leakoff and the #1 Seal Return Line is lowered causing #1 Seal Leak Return flow to Rise.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Incorrect. VCT pressure is used to provide back pressure to the RCP #1 Seal Return, As VCT pressure is lowered, the D/P between the #1 Seal Leakoff and the #1 Seal Return Line is lowered causing #1 Seal Leak Return flow to Rise.

C. Incorrect. When reducing VCT Pressure, the pressure reduction should be done very slowly, at any VCT pressure, to ensure gases do not come out of solution at the suction of the charging pumps. The 15 psig limit only applies to RCP operation.

D. Incorrect. When reducing VCT Pressure, the pressure reduction should be done very slowly, at any VCT pressure, to ensure gases do not come out of solution at the suction of the charging pumps. The 15 psig limit only applies to RCP operation.

Technical Reference(s): OTN-BG-00004, VCT Atmospheric Control, REV 11 References to be provided to applicants during examination: None Learning Objective: T61.0110, LP-11, Obj. AA. OTN-BG-00004, VCT Atmospheric Control, 1.

Explain the precautions and limitations and bases pertaining to Minimum and Maximum VCT Pressures.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 003 Reactor Coolant Pump Group # 1 K/A # 2.1.28 Importance Rating 4.1 Knowledge of the purpose and function of major system components and controls.

Question #29 Which of the following describes the purpose of the Reactor Coolant Pump (RCP) flywheel?

A. maintains flow after an RCP trip to prevent exceeding KW/FT limits B. prevent damage to the RCP anti-reverse rotation device during a pump shutdown C. maintain flow after an RCP trip for initial decay heat removal following a loss of site power D. allow the RCP to coast down slowly after tripping to prevent impact damage to the RCP seals Answer: C Explanation: Per the UFSAR, Sufficient pump rotation inertia is provided by a flywheel, in conjunction with the impeller and motor assembly, to provide adequate flow during coastdown. This forced flow following an assumed loss of pump power, and the subsequent natural circulation effect provides the core with adequate cooling flow. The UFSAR further states that It is important to reactor protection that the reactor coolant flow is maintained for a short time after a pump trip in order to remove heat stored in the fuel elements of the core. In order to provide this flow after interruption of power to the pump, each reactor coolant pump is provided with a flywheel. The rotating inertia of the pump, motor, and flywheel is employed during the coastdown period to continue the reactor coolant flow.

A. Incorrect - This is the function of the RTC Trip function #10 Reactor Coolant flow low. Per reference #3, The Reactor Coolant Flow - Low trip Function ensures that protection is provided against violating the DNBR limit due to low flow in one or more RCS loops, B. Incorrect - the antirotation device engages at low very low pump shaft RPMs, The flywheel would delay engagement of this device but not prevent damage to it.

C. Correct - see explanation above.

D. Incorrect - Plausible because the flywheel does allow the RCP to coast down slowly, however this is not to protect the seals. Starting the lift oil pump prior to starting the RCP helps prevent damage to the RCP bearings and RCP seals upon startup.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s):

(1) CALLAWAY UFSAR, Rev OL-20 11/13, Section 5.4.1.1 Reactor Coolant Pumps, Design st Basis. 1 paragraph. Page 5.4-1.

(2) CALLAWAY UFSAR, Rev OL-20 11/13, Section 5.4.1.3.2 Reactor Coolant Pumps, st Coastdown Capability, 1 paragraph. Page 5.4-4.

(3) Callaway Technical Specification Bases, Revision 11, for RTS Instrumentation - function 10, Reactor Coolant Flow - Low, page B 3.3.1-23.

References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems LP-09, BB_ RCS - Objective F DESCRIBE the purpose and operation of the following RCP components: #10 Flywheel Question Source: Bank # __X L15815____

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

CFR: 41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 004 Chemical and Volume Control Group # 1 K/A # A2.25 Importance Rating 3.8 Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Uncontrolled boration or dilution.

Question #30 The plant is operating at 100% power when VCT Level Channel BG LT-185 Fails LOW.

Which of the following describes:

(1) The plant response to this failure; And (2) The procedural action to be taken?

A. (1) VCT Outlet valve BG LCV-112B will close (2) Reduce Charging to RCP Seals Only B. (1) VCT Outlet valve BG LCV-112B will close (2) Place RCS Makeup Control selector switch in OFF C. (1) VCT Outlet valve BG LCV-112C will close (2) Reduce Charging to RCP Seals Only D. (1) VCT Outlet valve BG LCV-112C will close (2) Place RCS Makeup Control selector switch in OFF Answer: C Explanation:

A. Incorrect. BG LCV 112B received a close signal from VCT level channel BG LI-112, and would close if BG LI-112 failed low, but does not respond to a BG LI-185 failure. The action taken is correct for response to a failure which results in a VCT to RWST suction swap.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Incorrect. BG LCV 112B received a close signal from VCT level channel BG LI-112, and would close if BG LI-112 failed low, but does not respond to a BG LI-185 failure. The action listed is in response to a VCT level channel BG LI-149 failure, which provides input to normal makeup to the VCT and level divert functions.

C. Correct. VCT level channel BG LT-185 provides input to BG LCV-112C and BN LCV-112E.

Upon failing low, a suction swap over occurs for the charging pumps causing BG FCV-112C to Close and BN LCV-112E to open allowing an uncontrolled boration from the RWST. In response to this failure OTO-BG-00004 directs action to Reduce charging to seals only in order to minimize the amount of boron added to the RCS and the impact of the uncontrolled boration.

D. Incorrect. BG LCV-112C will close, however the action listed is in response to a VCT level channel BG LI-149 failure, which provides input to normal makeup to the VCT and level divert functions.

Technical Reference(s): OTO-BG-00004, VCT Level Channel Failures REV 18 References to be provided to applicants during examination: None.

Learning Objective: T61.003B LP-B-34, Obj. D. Given a set of plant conditions or parameters indicating a VCT level channel failure, Analyze the correct procedure to be utilized and the required actions to stabilize the plant.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 005 Residual Heat Removal Group # 1 K/A # A1.01 Importance Rating 3.5 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the RHRS controls including: Heatup/cooldown rates Question #31 Given the following plant conditions:

• The plant is in Mode 4 with A RHR Train in service.

Cooldown rate is currently 95F/hr and slowly rising, the operator is directed to REDUCE the cooldown rate.

In order to accomplish this, the operator will adjust the output of EJ HIC-606 controller such that EJ HV-606 ______(1)_____, and total RHR flow to the RCS will _____(2)____.

A. (1) closes (2) lower B. (1) closes (2) remain constant

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. (1) opens (2) rise D. (1) opens (2) remain constant Answer: B Explanation:

A. Incorrect. RHR heatup and cooldown rates are controlled with EJ HIC-606. By raising the output the valve will close allowing less flow through the RHR heat exchanger which in turn will lower the temperature out of the RHR system which will cause the cooldown rate to lower. The common misconception that the RHR Heat Exchanger bypass valve will not respond to the changes in system flow, therefore by raising EJ HIC-606, it would be incorrectly assumed that system flow will lower.

B. Correct. RHR heatup and cooldown rates are controlled with EJ HIC-606. By raising the output the valve will close allowing less flow through the RHR heat exchanger which in turn will raise the temperature out of the RHR system which will cause the cooldown rate to lower. With the controller for EJ FK-618 in Automatic control, the valve will reposition to maintain a constant return flow to the RCS.

C. Incorrect. By lowering the output of the valve controller EJ HIC-606, it will cause the valve to open which will allow the RCS cooldown rate to rise. Using the same misconception stated in explanation A, system flow would rise as the heat exchanger outlet valve EJ HIC-606 controller output is lowered.

D. Incorrect. The cooldown rate will rise as EJ HIC-606 output is lowered, however flow will remain constant if EJ FK-618 is controlling the bypass flow in AUTOMATIC.

Technical Reference(s): OTN-EJ-00001 Addendum 3, Placing A RHR Train In Service for RCS Cooldown, Rev 20 References to be provided to applicants during examination: None.

Learning Objective: T61.0110 Systems LP-07, Obj. J. Identify the RHR System Main Control Board controls, alarms, and indications and Describe how each is used to predict, monitor, or control changes in the RHR System.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

NRC Site-Specific Written Examination Callaway Plant Reactor Operator 10 CFR Part 55 Content:

(CFR: 41.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 006 Emergency Core Cooling Group # 1 K/A # K5.06 Importance Rating 3.5 Knowledge of the operational implications of the following concepts as they apply to ECCS: Relationship between ECCS flow and RCS pressure.

Question #32 Given the following plant conditions:

• RCS is at 380°F and 750 psig with a plant cooldown in progress

• A Loss of Coolant Accident occurs o RCS pressure is now 150 psig and lowering o Containment Pressure is 3.6 psig and slowly rising Which of the following describes the status of operating ECCS equipment?

A. SI Accumulator level is stable; RHR injection flow is rising.

B. SI Accumulator level is lowering; RHR injection flow is zero.

C. SI Accumulator level is stable; RHR injection flow is zero.

D. SI Accumulator level is lowering; RHR injection flow is rising.

Answer: A Explanation: During the cooldown Steam Line Pressure SI and RCS Pressure SI signals are blocked, however the containment pressure SI is still Active and will initiate an automatic SI at 3.5 psig.

A. Correct. Because the initial RCS Pressure is less than 1000 psig the SI Accumulators have been isolated with power removed from the discharge valves, so on the SI they do not respond and levels will remain stable. With RCS Pressure at 150 psig the RHR system will be injecting and as pressure continues to lower the discharge flow will rise.

B. Incorrect. Plausible if candidate believes that the SI accumulators are still aligned to the RCS and will discharge on an SI as pressure lowers. Also RHR flow is plausible to be at zero if the candidate believes that the RHR pumps will be discharging against the original pressure of 650 psig, which is greater than the shutoff head of the RHR pumps.

C. Incorrect. Plausible because SI Accumulators are Isolated and level will remain stable, and if the candidate assumes that RHR pumps are discharging against the original pressure of 650 psig

NRC Site-Specific Written Examination Callaway Plant Reactor Operator instead of the current value, they could assume that the RCS pressure is greater than the shutoff head of the RHR pumps.

D. Incorrect. Plausible if candidate believes that the SI accumulators are still aligned to the RCS and will discharge on an SI as pressure lowers. Also, the RHR system will respond to the lowering RCS pressure and will be discharging to the RCS with flow rising as RCS pressure is lowered.

Technical Reference(s): OTN-EP-00001, Addendum 6, SI Accumulator Isolation and Restoration, REV 1. , BD-E-1, Loss of Reactor Or Secondary Coolant Background Document, REV 9 References to be provided to applicants during examination:

Learning Objective: C DESCRIBE the phases of operation of the ECCS.

Question Source: Bank # ___L16242___

Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 007 Pressurizer Relief/Quench Tank Group # 1 K/A # A2.02 Importance Rating 2.6 Ability to (a) predict the impacts of the following malfunctions or operations on the P S; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Abnormal pressure in the PRT.

Question #33 Given the following plant conditions:

• The plant is at 100% power when the following alarm is received:

o 34E, PRT PRESS HI

• PRT parameters are as follows:

o pressure is 7 psig and RISING SLOWLY.

o level is 67% and STABLE.

What procedural action is directed to prevent PRT rupture disc operation for these conditions?

(1) The PRT rupture disc will actuate when pressure rises to ____.

And (2) To prevent PRT rupture disc operation the crew will ________.

A. (1) 50 psig (2) vent the PRT to a Gas Decay Tank to lower pressure B. (1) 100 psig (2) vent the PRT to a Gas Decay Tank to lower pressure C. (1) 50 psig (2) drain the PRT to the containment sump to reduce level and pressure D. (1) 100 psig (2) drain the PRT to the containment sump to reduce level and pressure Answer: B

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Explanation:

A. Incorrect. The PRT Rupture Disc is designed to rupture to relieve pressure between 86-100 psig in the PRT. In accordance with OTA-RK-00018 Add. 34E for PRT High Pressure, the high pressure should be addressed with OTN-BB-00004 to Vent the PRT to lower pressure.

B. Correct. The PRT Rupture Disc is designed to rupture to relieve pressure between 86-100 psig in the PRT. In accordance with OTA-RK-00018 Add. 34E for PRT High Pressure, the high pressure should be addressed with OTN-BB-00004 to Vent the PRT to lower pressure.

C. Incorrect. If the pressure increase is due to a level change, then draining the PRT is the correct method to lower level and pressure, however level is stable and therefore should be vented to maintain level while lowering pressure.

D. Incorrect. If the pressure increase is due to a level change, then draining the PRT is the correct method to lower level and pressure, however level is stable and therefore should be vented to maintain level while lowering pressure.

Technical Reference(s): OTN-BB-00004, Pressurizer Relief Tank, REV 37, OTA-RK-00018 Add 34E, REV 0 References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-09, Obj. U. Explain the precautions, limitations, and bases for the following processes/conditions associated with OTN-BB-00004, Pressurizer Relief Tank: 4.

PRT Venting.

Question Source: Bank # ___L16178___

Modified Bank # ______

New _______

Question History: Last NRC Exam _____Callaway 2005 ILT NRC Exam_______

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 008 Component Cooling Water Group # 1 K/A # A4.01 Importance Rating 3.3 Ability to manually operate and/or monitor in the control room: CCW indications and controls.

Question #34 Given the following plant conditions:

• The plant is operating at 100% power

• A CCW Pump is in service supplying the Service Loop

• A LOCA and Safety Injection occur 5 minutes later, the Reactor Operator performing E-0, Reactor Trip or Safety Injection, Attachment A observes the following indications:

• The AMBER light on EG HIS-22 for B CCW Pump is LIT

• Annunciator 53B, CCW PMP B/D TROUBLE, LIT

• Annunciator 53C, CCW PMP B FLOW LO, LIT

• Annunciator 54C, CCW PMP D FLOW LO, LIT Which of the following describes the CCW system line up following the completion of E-0, Reactor Trip or Safety Injection, Attachment A?

A. (1) A CCW Train is supplied by A CCW Pump ONLY (2) D CCW pump is MANUALLY started to supply B Train CCW B. (1) A CCW Train is supplied by A CCW Pump ONLY (2) D CCW pump is AUTOMATICALLY started to supply B Train CCW C. (1) A CCW Train is supplied by A and C CCW Pumps (2) D CCW pump is MANUALLY started to supply B Train CCW D. (1) A CCW Train is supplied by A and C CCW Pumps (2) D CCW pump is AUTOMATICALLY started to supply B Train CCW Answer: A Explanation:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator A. Correct. Based on the indications the candidate must conclude that the neither B or D CCW pumps are running. The D CCW pump will be manually started while performing attachment A.

The A CCW pump will continue to run supplying the A train CCW system which prevents the C CCW pump from starting.

B. Incorrect. Plausible because A Train CCW pump will continue to supply the A train, On the LOCA sequencer, D CCW pump should Automatically start if B fails to start, the annunciators for B/D pressure low is on the B Train CCW pump discharge header common to both B and D pumps, this is one of the Auto start signals to the D CCW pump, but is blocked with an SI signal present.

C. Incorrect. Plausible if the candidate assumes that all the pumps will receive a start signal on the SI, and that D failed to start. The A CCW pump will continue to run supplying the A train CCW system which prevents the C CCW pump from starting.

D. Incorrect. Plausible if the candidate assumes that all the pumps will receive a start signal on the SI, and that D CCW pump Automatically starts on the SI signal present if B CCW fails to start.

Technical Reference(s): Drawing J-22EG01B, C/D CCW pump logic diagram, REV 0, and Drawing J-22EG01A, A/B CCW pump logic diagram.

References to be provided to applicants during examination: None Learning Objective: T61.0110, LP-10, Obj. B. Draw, Label, and Explain a one line diagram of the CCW system under Normal, Safety Injection, and, Containment Isolation Phase B Conditions to include the following: 1a. CCW pumps.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _______N/A_____

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

K/A Match: Candidate must interpret the annunciators given to determine the status of the B Train CCW system and monitor the system response to ensure the proper system alignment is obtained during an SI condition.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 008 Component Cooling Water Group # 1 K/A # K1.05 Importance Rating 3.0 Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: Sources of makeup water.

Question #35 Which of the following is the EMERGENCY makeup supply to the Component Cooling Water System?

A. Fire Protection System B. Demineralized Water System C. Reactor Makeup Water System D. Essential Service Water System Answer: D Explanation:

A. Incorrect. The fire protection system is an emergency makeup source that can be used to restore other systems such as the spent fuel pool, but is not a source for the makeup of CCW.

B. Incorrect. Normal Makeup water is supplied to the CCW surge tank by Demineralized water, however if the leak exceeds the capacity of Demineralized Water then the ESW system can be used as an emergency makeup source.

C. Incorrect. The reactor makeup water system is a subsystem of the Demineralized water system used primarily for makeup to the VCT for RCS volume control. It is a clean source of water compared to ESW making it a plausible choice for makeup.

D. Correct. Emergency Makeup water can be supplied directly to the pump suction header by the ESW system via motor operated valves that are operated from the control room.

Technical Reference(s): OTO-EG-00001, CCW System Malfunction, REV 14 References to be provided to applicants during examination: None Learning Objective: T-61.0110, LP-10, Obj. E. List the systems that interface with the CCW system and Explain how a loss of the interfacing system or CCW will affect the other.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Source: Bank # ___L16321___

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.8)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 010 Pressurizer Pressure Control Group # 1 K/A # K2.03 Importance Rating 2.8 Knowledge of bus power supplies to the following: Indicator for PORV position Question #36 During a panel walkdown, a Reactor Operator (RO) notices that NO indicating lights are LIT on BB-HIS-455A, the A PZR PORV hand indicating switch.

To check for a loss of power, the RO should dispatch the operations technician to __________.

A. NK01 B. NK02 C. NK03 D. NK04 Answer: A Explanation: the A PZR PORV indication and control is powered from NK01.

A. Correct B. Incorrect - NK02 is a safety related DC bus but does not support the A PZR PORV C. Incorrect - NK03 is a safety related DC bus but does not support the A PZR PORV D. Incorrect - the B PZR PORV is supplied from NK04 Technical Reference(s): OTO-NK-00002, Loss of Vital 125 VDC Bus, Revision 13. Attachment L page 2 of 5. (Page 36 of overall procedure.)

References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems LP-09, BB_ RCS - B. DESCRIBE the purpose and operation of the following RCS components to include interlocks, controller operations and power supply: #5 Power Operated Relief Valves (PORVs)

Question Source: Bank # ______

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

CFR: 41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 012 Reactor Protection Group # 1 K/A # A1.01 Importance Rating 2.9 Ability to predict and/or monitor Changes in parameters (to prevent exceeding design limits) associated with operating the RPS controls including: Trip setpoint adjustment .

Question #37 Given the following plant conditions:

• The plant is operating at 100% power.

• Pressurizer Pressure is currently 2215 psig.

• The Reactor Operator is directed to take manual control of the Master Pressurizer Pressure Controller and restore pressure to 2235 psig.

As Pressurizer Pressure is raised which of the following correctly describes the response of the OTT and OPT Reactor Trip setpoints?

OTT OPT A. rises rises B. rises stays the same C. stays the same rises D. stays the same stays the same Answer: B Explanation:

A. Incorrect. See explanation B.

B. Correct. OTT setpoint is changed by three variables; Tavg, Pressurizer Pressure, and AFD.

OPT setpoint is changed by two variables; Tavg, and the rate of change of RCS Tavg.

Changing the Pressure of the Pressurizer and RCS alone will only cause the OTT setpoint to change. Based on the equation used to calculate the setpoint in the T/S bases this will cause the OTT setpoint to rise.

C. Incorrect. See explanation B.

D. Incorrect. See explanation B.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): T/S 3.3.1 Reactor Trip Instrumentation, and T/S 3.3.1 Bases.

References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-27 Obj. C. List all the Reactor Trip Signas supplied to RPS including setpoint, coincidence, interlocks and protection afforded. Obj. J. State the Limiting Conditions for Operation and Bases associated with the RPS related Technical specifications.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 012 Reactor Protection Group # 1 K/A # K3.03 Importance Rating 3.1 Knowledge of the effect that a loss or malfunction of the RPS will have on the following:

SDS.

Question #38 Given the following plant conditions:

• The plant was operating at 100% power when a Reactor Trip Occurred

• A Reactor Trip Breaker (RTA) indicates OPEN

• B Reactor Trip Breaker (RTB) indicates CLOSED Based on these conditions, at what temperature will the condenser steam dumps attempt to maintain RCS Temperature?

A. 550F B. 557F C. 559F D. 561F Answer: C Explanation:

A. Incorrect. This is a temperature associated with the steam dump controller for the P-12 interlock, which allows continued cooldown only on the group 1 valves once 550F is reached.

B. Incorrect. Under normal conditions following a reactor trip, the B Trip breaker being open will make up the P-4 contact allowing the steam dumps to be controlled by the Reactor Trip controller which will modulate steam dumps to control at 557F. This is only driven off of the B RTB, with this breaker still closed, the circuit does not recognize the reactor trip and will not transfer to this controller.

C. Correct. With the B Reactor Trip Breaker still closed, the P-4 contact which would activate the Reactor Trip controller for steam dumps would not be met, and the controller will continue to operate on the Load Reject Controller. The Tave Load Reject Controller has a 2F dead band on the lower end for normal operations to allow rod motion without operation of the steam dumps, however this correlates to a lower temperature of 559F at which the controller will attempt to maintain temperature with a Tref signal of zero.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator D. Incorrect. 561F correlates to the saturation temperature of the steam generators at the lift setpoint for the S/G Atmospheric Steam Dumps. If the condenser steam dump system was not available, the S/G ASDs will cycle to maintain RCS Tavg at approximately 561F.

Technical Reference(s): Drawing 7250 D64 Sheet 10, Steam Dump System Logic Operation References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-20 Obj. I. Discuss the four Steam Dump permissive interlocks and Explain the effects of each on system operations.

Question Source: Bank # ___L13070___

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 013 Engineered Safety Features Actuation Group # 1 K/A # K6.01 Importance Rating 2.7 Knowledge of the effect of a loss or malfunction on the following will have on the ESFAS:

Sensors and detectors.

Question #39 Given the following plant conditions:

• Containment Pressure Transmitter GN PT-935B has failed LOW.

• NO Technical Specification actions have been taken.

What is the resultant coincidence required for the actuation of a Containment Spray signal based on Containment pressure?

A. 1 out of 2 B. 1 out of 3 C. 2 out of 2 D. 2 out of 3 Answer: D Explanation:

A. Incorrect. Containment Pressure has 3 channels that feed into the Safety Injection Actuation Circuit. Tech Spec Actions for a failed channel place the channel in Trip. This leaves two channels available to actuate SI on high containment pressure. Because one channel is already in the Trip condition, only 1 channel of the remaining 2 will initiate a Safety Injection.

B. Incorrect. Containment Pressure is monitored using 4 channels, 2/4 channels of High Containment Pressure will initiate a Containment Spray Actuation Signal. Therefore if it is incorrectly assumed that the low failure will actuate the bistable, or they may assume that T/S actions will place the channel in TRIP such as it will with the Safety Injection signal resulting in a 1 out of 3 logic.

C. Incorrect. If it is assumed that Containment Pressure is only monitored by 3 channels and the candidate knows that a low failure will NOT actuate the bistable, then the remaining coincidence will be 2 out of 2 remaining channels.

D. Correct. The containment spray system is actuated by 2 out of 4 containment pressure channels indicating HIGH, if a channel fails low with no action taken, the resultant coincidence will be 2 out of the remaining 3 channels.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): E-0, Reactor Trip or Safety Injection, REV 16, T/S 3.3.2.

References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-17, Obj. D. State the conditions that will initiate a Safety Injection Signal and Describe the conditions necessary to reset the signal.

Question Source: Bank # ___ ___

Modified Bank # ___ L7244___

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 022 Containment Cooling Group # 1 K/A # A4.01 Importance Rating 3.2 Ability to manually operate and/or monitor in the control room: CCS fans.

Question #40 Given the following plant conditions:

• RCS Temperature is 250 deg F and STABLE.

• 2 CRDM Fans are in service.

• CRDM shroud temperature is 150 deg F and STABLE.

• The crew receives a high vibration alarm on Computer Point GNY0019, CTMT CLR FAN D VIB, that indicates excessively high vibration on the "D" Containment Cooler.

• Currently, all containment coolers are in operation in "SLOW" speed.

• Containment temperature is 110 deg F STABLE.

What action is required?

A. Stop the "D" Containment Cooler and shift all others to "FAST" speed.

B. Stop the "D" Containment Cooler. No other actions are required.

C. Stop the "D" Containment Cooler and verify all CRDM fans are running.

D. Stop the "D" Containment Cooler and ensure the Pressurizer Cooling Fan is running.

Answer: D Explanation: Per OTN-GN-00001 Section 5.18 Containment Cooler Fan Malfunction if CTMT COOLER UNIT D, is tripped and RCS Temp is greater than 120°F, ENSURE the Pressurizer Cooling Fan is running per Section 5.10.

A. Incorrect - see above step 5.2.5 OPERATE containment coolers as required to maintain containment air temperature between 50°F and 120°F. is met and there is no need to shift fans to fast speed.

B. Incorrect - see above - the pressurizer fan is required to be in service.

C. Incorrect - see above - per section 5.7 2 CRDM are required to be in service and are. The temperature limit is 165F and indications given say no additional CRDM fans are required.

D. Correct - see above

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): OTN-GN-00001, Rev 27 Section 5.18.2 References to be provided to applicants during examination: None Learning Objective: B DESCRIBE the purpose and operation of the following containment cooling system components.

1.Containment Fan Coolers 2.Hydrogen Mixing Fans 3.Cavity Cooling Fans 4.CRDM Cooling Fans 5.Pressurizer Cooling Fan 6.Elevator Machine Room Exhaust Fan Question Source: Bank # __L16565____

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.7 / 45.5 to 45.8)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 026 Containment Spray Group # 1 K/A # K4.04 Importance Rating 3.7 Knowledge of CSS design feature(s) and/or interlock(s) which provide for the following:

Reduction of temperature and pressure in containment after a LOCA by condensing steam, to reduce radiological hazard, and protect equipment from corrosion damage (spray).

Question #41 Which of the following completes the statement below?

Following a LOCA or Steam Line Break, the Containment Spray System is designed to ______(1)_______, and to maintain containment pressure below the MAXIMUM containment design pressure of (2) .

A. (1) minimize offsite radiation levels (2) 47.8 psig B. (1) limit containment temperature to 120 F (2) 47.8 psig C. (1) minimize offsite radiation levels (2) 60 psig D. (1) limit containment temperature to 120 F (2) 60 psig Answer: C Explanation:

A. Incorrect. The maximum expected value for the peak containment pressure per the safety analysis during a design basis LOCA is calculated to be 47.8 psig, however the Maximum design limit for containment is 60 psig.

B. Incorrect. See A., the design basis for the containment cooling system during normal operation is to maintain ambient containment temperature to less than 120F. This system is credited during and DBA to aid in reducing containment temperature and thus pressure to reduce offsite rad levels. However containment spray is not designed to maintain a specific temperature.

C. Correct. The Containment Spray system is designed to maintain the pressure in containment less than the design limit of 60 psig, and to minimize offsite radiation levels.

D. Incorrect. See B.

Technical Reference(s): FSAR Chapter 6, Section 6.2.2.1, and Technical Specification Bases 3.6.4.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination: None Learning Objective: T61.0110, LP-18, Obj. A. State the function and Explain the design criteria of the Containment Spray System.

Question Source: Bank # __L16800____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 039 Main and Reheat Steam Group # 1 K/A # K5.08 Importance Rating 3.6 Knowledge of the operational implications of the following concepts as the apply to the MRSS: Effect of steam removal on reactivity.

Question #42 Given the following plant conditions:

• The crew is performing a startup following a Middle of Life reactor trip

• Reactor power is at 10E-8 amps

• B Steam Generator ASD Fails OPEN

• Tavg stabilizes at 548F In response to this event, reactor power will stabilize ____(1)____ the Point of Adding Heat, and to comply with Technical Specifications, the crew must place the reactor in a subcritical condition within ____(2)____.

(1) (2)

A. at 30 minutes B. at 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> C. above 30 minutes D. above 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> Answer: C Explanation:

S/G ASDs are rated for approximately 5% steam flow per valve, therefore if a PORV fails open it would cause the reactor to stabilize at approximately 5% power, above the POAH (1-3%).

A. Incorrect. Power will remain above the POAH providing negative reactivity feedback due to heating up the RCS.

B. Incorrect. Power will remain above the POAH providing negative reactivity feedback due to heating up the RCS. 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> is incorrect.

C. Correct. The ASD failing open will cause Tavg to lower resulting in a positive reactivity addition to the reactor. Power will rise above the POAH until the negative reactivity from the temperature feedback is provided due to the rising Tavg, at which point power will stabilize at a

NRC Site-Specific Written Examination Callaway Plant Reactor Operator value corresponding to the steam demand from the open ASD. Tavg is below T/S 3.4.2 minimum temperature for criticality of 551F, and the plant must be placed in Mode 2 with Keff <1 within 30 minutes.

D. Incorrect. See C for POAH explanation. T/S action is 30 minutes not 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Technical Reference(s): T/S 3.4.2, RCS Minimum temperature for Criticality.

References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-09 Obj. K. State the LCOs for the RCS minimum temperature for criticality technical specification and Identify the RCS instruments that these technical specifications are based on.

T61.0110 LP-20 Obj. N. Explain the following precautions, limitations and bases associated with OTN-AB-00001, Main Steam and Steam Dump System; 2. Approach to criticality.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000007 (BW/E02&E10; CE/E02) Reactor Trip - Group # 1 Stabilization - Recovery / 1 K/A # EA1.04 Importance Rating 3.6 Ability to operate and monitor the following as they apply to a reactor trip: RCP operation and flow rates.

Question # 43 Given the following conditions:

• SI actuated due to a LOCA.

• "A" and "B" SI Pumps are TRIPPED.

• Boron Injection header flow indicates 425 GPM on both trains.

• RCS pressure is 1250 psig.

• RCS Flow is 104% for each Loop.

• Containment pressure is 5 psig.

• The crew is performing actions of E-0, Reactor Trip or Safety Injection.

Which of the following describes the required action and reason for the action with respect to the Reactor Coolant Pumps (RCP's)?

A. Stop all RCP's to minimize fluid mass loss out of the break.

B. Stop all RCP's to minimize secondary side inventory depletion.

C. Leave all RCP's running to provide forced cooling flow of the RCS.

D. Leave all RCP's running to prevent phase separation of RCS liquid.

Answer: A Explanation:

A. Correct. For small break LOCA, RCPs are tripped based on an RCS pressure of less than 1425 psig with at least one high head ECCS pump injecting (Met with BIT flow) in order to minimize the RCS inventory loss through the break.

B. Incorrect. Stopping all RCPs is correct however the reason stated is incorrect for a small break LOCA, but is correct for a loss of secondary heat sink.

C. Incorrect. RCPs should be secured, however RCPs will be left running if conditions are met for Degraded Core Cooling in which operators are directed to leave RCPs running to maintain forced flow to the Core for cooling.

D. Incorrect. RCPs should be secured, During the analysis of a Small Break LOCA RCPs are

NRC Site-Specific Written Examination Callaway Plant Reactor Operator left running in the initial phases of the accident to maintain adequate mixed flow through the operational loops to maintain core cooling and minimize break flow until the point where a predetermined setpoint (<1425 psig in the RCS) is reached where it is determined that the RCS is at saturation and steam voids will occur in the steam generator u-tubes when RCPs are secured.

Technical Reference(s): ERG Executive Volume, Background Information for WOG Emergency Response Guidelines Generic Issue RCP Trip/Restart REV 3.

References to be provided to applicants during examination: None Learning Objective: T61.003D, LP-D-04, Obj. G. Describe the Criteria and the Basis for information as stated on the E-0, Reactor Trip Or Safety Injection, Foldout Page.

Question Source: Bank # ______

Modified Bank # ___L16207___

New _______

Question History: Last NRC Exam ______Callaway 2005 ILT NRC Exam______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000009 Small Break LOCA / 3 Group # 1 K/A # EK3.15 Importance Rating 3.2 Knowledge of the reasons for the following responses as the apply to the small break LOCA: Closing of RCP thermal barrier outlet valves.

Question # 44 Given the following plant conditions:

• Reactor power is 100%

• 'A' Train Component Cooling Water is in service Subsequently:

• Annun 61B, PROCESS RAD HI, alarms

• Annun 51D, CCW SRG TK A LEV HILO, alarms

• 'A' CCW Surge Tank level is 91%

• PZR Level is 56% and lowering slowly Which of the following has caused the above indications and why is this component isolated?

A. Seal Water Heat Exchanger leak Prevent contamination of the CCW system B. RCP 'D' Thermal Barrier Heat Prevent contamination of the Exchanger leak CCW system C. Seal Water Heat Exchanger leak Prevent an increase in Reactor Power D. RCP 'D' Thermal Barrier Heat Prevent an increase in Reactor Exchanger leak Power Answer: B Explanation: The CCW surge tank hi level alarm is a 87 inches (85.4%)

NRC Site-Specific Written Examination Callaway Plant Reactor Operator A. Incorrect - A leak in the Seal Water heat exchanger would case CCW surge tank to lower. The conditions given indicate a high level in the CCW surge tank. A leak in the seal water HX would cause a flow of water from CCW into the reactor coolant causing a dilution and increase in reactor power NOT contamination of the CCW system.

B. Correct - A leak in the RCP D thermal barrier HX would cause CCW surge tank level to increase. This matches the condition given. A leak in the RCP D thermal barrier HX would cause contamination of the CCW system.

C. Incorrect - A leak in the Seal Water heat exchanger would case CCW surge tank to lower. The conditions given indicate a high level in the CCW surge tank. A leak in the seal water HX would cause a flow of water from CCW into the reactor coolant causing a dilution and increase in reactor power D. Incorrect - A leak in the RCP D thermal barrier HX would cause CCW surge tank level to increase. This matches the condition given. But A leak in the RCP D thermal barrier HX would cause contamination of the CCW system and NOT cause an increase in reactor power.

Technical Reference(s): E-1 and OTO-EG-00001, CCW System Malfunction References to be provided to applicants during examination: None Learning Objective: H DESCRIBE the plant response and DISCUSS any possible detrimental affects of leaks in the following Heat Exchangers (HXs):

1. Letdown HX

2. Seal Water HX

3. RHR HX

4. Fuel Pool Cooling HX

5. CCW HX

6. RCP Thermal Barrier HX Question Source: Bank # ______

Modified Bank # ___L16616___

New _______

Question History: Last NRC Exam __Modified 2013__________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR 41.10)

Comments: The distractors are plausible because the seal water heat exchanger is cooled by CCW. The candidate must be able to differentiate the indication of a leak in the correct heat exchanger. Reactor power would increase if the leak was in the seal water heat exchanger.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Group # 1 000011 Large Break LOCA / 3 K/A # EK2.02 Importance Rating 2.6 Knowledge of the interrelations between the Large Break Loca and the following: Pumps.

Question # 45 Given the following conditions:

• The plant was operating at 100% power when a LOCA occurred.

• The crew has just transitioned from E-1 to ES-1.2, Post LOCA Cooldown and Depressurization, and currently performing step 1 of ES-1.2.

Current plant conditions:

Core Exit Thermocouples 375°F - Lowering Wide Range Thot 425°F - Lowering RCS Pressure 400 PSIG and Stable Subcooling Meter 3°F -Subcooled Containment Pressure 28 PSIG - Lowering Containment Temperature 175°F - Stable Containment Radiation 5 x 103 R/HR - Stable RWST Level 87% - Lowering Secondary Radiation Levels Normal S/G Levels 3.6% NR rising S/G Pressures Stable What is the status of the Reactor Coolant and RHR Pumps?

RCPs RHR Pumps Running Running A. NO YES B. YES NO C. YES YES D. NO NO

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Answer: D Explanation: Based on the given conditions, RCS pressure less than 1425 psig and Containment pressure greater than 27 psig, RCP should be secured. This is covered by both foldout page criteria and a step in E-0 prior to the transition to E-1. Based on the conditions given RHR pumps are secured due to SI pumps running and maintaining RCS pressure greater than 325 psig.

A. Incorrect - See above B. Incorrect - See above C. Incorrect - See above D. Correct - See above Technical Reference(s):

1. E-0 (and BD-E0) Reactor Trip or Safety Injection, Rev 16

2. E-1 Loss of Reactor or Secondary Coolant, Rev 17 References to be provided to applicants during examination: None Learning Objective: T61.003D LP-D-08 Obj I. STATE and EXPLAIN the parameters which are evaluated, including their Criteria and Basis, to transition from E-1 to other procedures.

Question Source: Bank # __R11859____

Modified Bank # ______

New _______

Question History: Last NRC Exam ___N/.A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR 41.8)

Comments:

E-1 also contains the same foldout page criteria as E-0 with respect to tripping RCPs

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000022 Loss of Rx Coolant Makeup / 2 Group # 1 K/A # AA2.04 Importance Rating 2.9 Ability to determine and interpret the following as they apply to the Loss of Reactor Coolant Makeup: How long PZR level can be maintained within limits.

Question # 46 Given the following conditions:

• The plant is at 100% power and at normal operating temperature, pressure, and pressurizer level

• BG FCV-124, Charging Flow Control Valve, malfunctions to limit charging flow to 55 gpm

• Normal letdown is in service at 75 gpm

• Identified leakage is at the Tech Spec limit

• RCP seal leak-off is 3 gpm per pump If NO operator action is taken, approximately how much time will elapse before the pressurizer heaters trip?

(Assume PZR level is 62 gal/%)

A. 39 minutes B. 58 minutes C. 74 minutes D. 81 minutes Answer: B Explanation:

A. Incorrect. If the 20 gpm difference between charging and letdown, with the correct identified leakage rate of 10 gpm used, a time of 39 minutes is calculated for reaching 17% PZR level B. Correct. Calculating a net loss from the RCS at 42 gpm,(12 gpm for RCP leakoff, 10 gpm leakage at TS limit for identified leakage, and 20 gpm charging letdown difference),Current PZR level at 100% of 56%, and PZR heater trip setpoint of 17%, results in 58 minutes elapsing until the PZR heaters trip.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Incorrect. If it is assumed incorrectly that identified TS limit is 1 gpm, the answer is calculated at 74 minutes.

D. Incorrect. If calculated net loss from the RCS is calculated at 30 gpm (10 identified leakage and 20 difference from charging and letdown) disregarding the RCP seal leakoff the time is 81 mintues.

Technical Reference(s): OSP-BB-00009 ADD1, REV 2, RCS Inventory Balance Excessive Leakage or Manual Calculation.

References to be provided to applicants during examination: None Learning Objective: T61.003B, LP-B12, Obj. D. Given a set of plant conditions or parameters indicating excessive Reactor Coolant leakage, ANALYZE the correct procedure(s) to be utilized and the required actions to stabilize the plant.

Question Source: Bank # ___L15796___

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR 43.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000025 Loss of RHR System / 4 Group # 1 K/A # AK1.01 Importance Rating 3.9 Knowledge of the operational implications of the following concepts as they apply to Loss of Residual Heat Removal System: Loss of RHRS during all modes of operation Question # 47 Given the following conditions:

• Plant is in Mode 5 following a shutdown for refueling.

• Mid-loop operations in progress.

• S/G hot and cold leg manways removed.

• S/G nozzle dams installed on hot legs.

• S/G nozzle dams NOT installed on cold legs.

• No other vents are open in the RCS.

• Loss of RHR cooling occurs.

Which of the following will occur as a result of this event?

Steam formation in the upper head will increase pressure A. and cause the PZR to refill rapidly.

B. enough to blow out the hot leg nozzle dams.

C. and displace water out the S/G hot leg manways.

D. and displace water out the S/G cold leg manways.

Answer: D Explanation:

A. Incorrect - Due to the elevation difference of the pressurizer and the vessel, the head of the water will not flow up into the pressurizer but will instead fill all the hot and cold legs first, since there is nothing to provide enough back pressure from the legs to force water into the pressurizer, this will not happen. It is plausible if the candidate does not understand the physical layout of the RCS with elevation relations for the Vessel/Legs/Pressurizer.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Incorrect - If all Nozzle dams were installed this is a possible correct answer, however with the cold leg nozzle dams NOT installed, the pressure will be relieved through the cold legs and S/G manways preventing the pressure rise from displacing the nozzle dams.

C. Incorrect - The water displaced water from the steam formation will flow into the hot legs, but will not exit the manways with the Hotleg nozzle dams installed.

D. Correct - With the loss of RHR cooling, the long term effect of not restoring cooling will be the formation of a steam bubble in the Reactor head causing a rising pressure in the RCS system forcing coolant out of the vessel into the hot and cold legs. With the S/G nozzle dams on the Cold legs not installed, this expansion of coolant will follow the path of least resistance and exit the RCS through the S/G manways.

Technical Reference(s): OTN-BB-00002 Addendum 7, Raising RCS Level to 6 inches below the RX vessel Flange, REV 2 References to be provided to applicants during examination: None Learning Objective: A State the significance of a loss of RHR(shutdown) cooling during reduced inventory operations.

Question Source: Bank # _L15232_____

Modified Bank # ______

New _______

Question History: Last NRC Exam _2005___________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR 41.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000026 Loss of Component Cooling Group # 1 Water / 8 K/A # AA1.01 Importance Rating 3.1 Ability to operate and / or monitor the following as they apply to the Loss of Component Cooling Water: CCW temperature indications.

Question # 48 The plant is operating at 100% power.

The following events occur:

• CCW flow is lost to the B RCP.

• Motor bearing temperature 175ºF and rising at 4ºF/minute.

• Stator Winding temperature 290ºF and rising at 3ºF/minute.

• Seal injection flow has been maintained to the RCP.

Which of the following describes the MAXIMUM time allowed before the crew must stop the B RCP?

A. 3 minutes B. 5 minutes C. 7 minutes D. 10 minutes Answer: B Explanation: Motor Bearing Temperature limit is 195°F. Motor Stator Winding Temperature limit is 311°F 195-175=20°. 20/4 = 5 minutes until temperature limit for motor bearing is reached.

311-290=21. 21/3 = 7 minutes until temperature limit for stator winding is reached.

A. Incorrect see explanation above. Plausible if student does not remember 195 and applies the 185 bearing #

B. Correct see explanation above This is the shortest time in which the RCP is required to be secured.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Incorrect see explanation above. Will this # is correct it is not the earliest time in which the RCP must be secured D. Incorrect see explanation above. Plausible because Attachment C step 1 says that if CCW is lost for 10 minutes to the RCP motor it must be secured.

Technical Reference(s): OTO-BB-00002, RCP Off-Normal, Rev 31 References to be provided to applicants during examination: None Learning Objective: G. Using OTO-BB-00002 RECOGNIZE the conditions that would require a Reactor Trip/Turbine Trip Question Source: Bank # __L15224____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____2005 NRC exam________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR 41.7)

Comments:

Match justification: A loss of CCW is provided in the question, and rising temperature values are given. Knowledge is required to answer the question of how to operate as a result of the temperatures, monitor the temperatures, and at which temperatures a reactor trip is required).

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000027 Pressurizer Pressure Control Group # 1 System Malfunction / 3 K/A # AA2.03 Importance Rating 3.3 Ability to determine and interpret the following as they apply to the Pressurizer Pressure Control Malfunctions: Effects of RCS pressure changes on key components in plant.

Question # 49 Given the following plant conditions:

• The plant is operating at 100% power.

• Pressurizer Pressure Control Instrument BB PT-457 is selected as the controlling Pressurizer Pressure channel.

• BB PT-457 Fails low.

Assuming NO Operator Action, which of the following describes the appropriate response of the pressurizer PORVs BB PCV-455A and BB PCV-456A?

A. ONLY BB PCV-455A will Open B. ONLY BB PCV-456A will Open C. BOTH BB PCV-455A and BB PCV-456A will Open D. NEITHER BB PCV-455A or BB PCV-456A will Open Answer: C Explanation:

A. Incorrect. When operating under Cold Overpressure Protection (COMS) each PORV is provided Train specific inputs to open. If the operator assumes the PORVs are Train specific and the assumption that instrument BB PT-457 feeds the opposite PORV this answer could be selected.

B. Incorrect. When operating under Cold Overpressure Protection (COMS) each PORV is provided Train specific inputs to open. If the operator assumes the PORVs are Train specific and the assumption that instrument BB PT-457 feeds the opposite PORV this answer could be selected.

C. Correct. PZR PORVs are automatically opened when 2/4 pressure instruments are greater than 2335 psig and operating independent of the Master Pressure Controller and are not train specific during normal operation. With the controlling instrument for PZR pressure failing low, the

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Master Pressure Controller responds to RAISE RCS pressure by energizing heaters and closing spray valves. Because no operator action is taken the input to the Master Pressure controller is never seeing the rising pressure and therefore spray valves never open to restore pressure.

D. Incorrect. If the candidate assumes that the indicated pressure, low, is how the actual plant pressure would trend following the failure of BB PT-457, then the operator could assume that plant pressure would never reach a point that would open both PORVs. The candidate could also assume that the plant response would be that of an actual HIGH failure. IF this was the case, the overall RCS pressure would begin to lower due to the master pressure controller response until a reactor trip and safety injection occurred. Pressure would not rise to a point when PORVs would open until late in the transient following a Safety Injection when the PZR level rises to a point to fill the PZR at which time BOTH PORVs would open.

Technical Reference(s): Functional Diagram Pressurizer Trip Signals Drawing 7250D64 Sheets 6, 17, and 18.

References to be provided to applicants during examination: None Learning Objective: T61.003B LP-B-41, Obj. C. Given a set of plant conditions or parameters indicating a Pressurizer Pressure Control Malfunction, Analyze the correct procedure to be utilized and the required actions to stabilize the plant.

Question Source: Bank # ___L15658___

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 45.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 000029 ATWS / 1 Group # 1 K/A # EK1.01 Importance Rating 2.8 Knowledge of the operational implications of the following concepts as they apply to the ATWS: Reactor nucleonics and thermo-hydraulics behavior.

Question # 50 Given the following plant conditions:

• The plant is operating at 100% power middle of life

• The Main Turbine tripped on a loss of condenser vacuum

• The Reactor failed to trip automatically With NO operator action, INITIALLY Steam Generator Water Level will

____(1)_____ and Reactor Power will ____(2)_____.

A. (1) rise (2) rise B. (1) rise (2) lower C. (1) lower (2) rise D. (1) lower (2) lower Answer: D Explanation: With the plant operating at 100% power, greater than P-9 (50%) power a turbine trip should result in an automatic reactor trip. When the turbine is tripped the reduction in steam load will result in steam generator pressure rising and a shrink effect in the steam generators causing level to lower. This rise in pressure will result in Tcold temperatures rising and due to the Moderator Temperature Coefficient providing negative reactivity, this will lower reactor power. In addition to the MTC providing negative reactivity, the large difference in Tavg/Tref after the turbine is tripped, will result in a rod motion demand to insert rods at the maximum speed 72 steps/min, which will also reduce reactor power.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator A. Incorrect. Plausible because the candidate could conclude that with no turbine or steam dumps available due to the loss of condenser vacuum, that there would be no reason for steam generator level to lower, and could assume that the reactor power will rise as temperature rises in the RCS. Temperature will rise, but the MTC will add negative reactivity as the temperature rises.

B. Incorrect. See Explanation and A C. Incorrect. See Explanation and A D. Correct. - See Explanation Technical Reference(s): FR-S.1, Response to Nuclear Power Generation/ATWS References to be provided to applicants during examination: None Learning Objective: Lesson Plan D-29, Obj B. Describe the Symptoms and/or Entry Conditions for: FR-S.1, Response to Nuclear power Generation/ATWS Question Source: Bank # ______

Modified Bank # _ L16783_____

New _______

Question History: Last NRC Exam ______Modified from Callaway 2011 NRC Exam______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR 41.10)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 073 Process Radiation Monitoring Group # 1 K/A # K1.01 Importance Rating 3.6 Knowledge of the physical connections and/or cause-effect relationships between the PRM system and the following systems: Those systems served by PRMs.

Question #51 Given the following plant conditions:

• The plant is operating at 100% power.

• A Hi Hi alarm is received on BM-RE-52, Steam Generator Blowdown Discharge Pumps Discharge Radiation Monitor Which of the following valves will automatically close?

A. BM-HV-1, S/G 'A' Blowdown Isolation Valve B. BM-HV-20, S/G 'B' Blowdown Upper Sample Isolation Valve C. BM-HV-37, S/G 'C' Blowdown Lower Sample Isolation Valve D. BM-HV-68, S/G 'D' Blowdown Outer Containment Sample Isolation Valve Answer: A Explanation:

A. Correct. Upon receipt of a HI HI Alarm, the rad monitor BM-RE-52 generates a BSPIS (Steam Generator Blowdown and Sample Process Isolation Signal). This signal in turn will close the Outer containment Isolation valves BM-HV-5,6,7,8, and Blowdown Isolation valves BM-HV-1,2,3,4.

B. Incorrect. Plausible because this valve does receive an automatic isolation signal. The Radiation monitor only actuates a BSPIS, a Separate Signal SGBSIS (Steam Generator Blowdown System Isolation Signal) is generated by a Safety Injection, Undervoltage on NB01 or NB02, or a Motor Driven AFW pump Actuation Signal (AFAS). This signal will close Lower Sample Isolation Valves, BM-HV-35,36,37,38, Upper Sample Isolation Valves, BM-HV-19,20,21,22, and Outer Containment Isolation Valves BM-HV-65,66,67,68.

C. Incorrect. See explanation B.

D. Incorrect. See explanation B.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): Drawing J-22BM02, Blowdown & Sample Process Isolation Signal. J-22BM04, Steam Generator Blowdown Isolation Valves.

References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-12, Obj. D. State the parameter that actuates the following signals, and Explain the actions that result from these signals; 2. Blowdown and Sample Process Isolation Signal (BSPIS).

Question Source: Bank # ______

Modified Bank # ___L4647___

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.5)

Comments:

K/A Match: Candidate must have knowledge of the S/G Blowdown Radiation Monitor interlocks associated with the valves in the S/G Blowdown system and how each responds to a HI HI alarm on the rad monitor.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 076 Service Water Group # 1 K/A # A3.02 Importance Rating 3.7 Ability to monitor automatic operation of the SWS, including: Emergency heat loads.

Question #52 Given the following plant conditions:

• Callaway is operating at 80% power.

• A leak in the Pressurizer Vapor Space results in the following parameters:

o Pressurizer Level: 60% and RISING o Pressurizer Pressure: 1830 psig and LOWERING o Containment Temperature: 130F and RISING o Containment Pressure: 0.5 psig and RISING o NR Steam Generator Levels: All 15% and LOWERING o Steam Generator Pressures: All 1100 psig and LOWERING Which of the following describes the IMMEDIATE response of the Essential Service Water System to this event?

A. The UHS cooling tower bypass valve closes B. The containment cooler bypass valve EF-HV-49 closes.

C. The component cooling water heat exchanger inlet valve EF-HV-51 opens.

D. The component cooling water heat exchanger outlet valve EF-HV-59 opens.

Answer: C Explanation:

The Pressurizer Pressure is below the SI setpoint of 1849 psig, therefore the ESW system will realign based on a Safety Injection. A reactor trip has occurred due to a valid SI signal.

A. Incorrect. Plausible because on a Safety Injection the ESW system aligns to discharge to the UHS cooling tower, however the UHS cooling tower bypass valve does not reposition on the SI signal alone, and will only close if ESW return temperature exceeds 84F and the ESW pumps have been running for 5 minutes. Incorrect because it is not an IMMEDIATE response of the ESW system.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Incorrect. Plausible because the containment cooler bypass valve does receive an automatic reposition signal on a Safety Injection to OPEN, not Close.

C. Correct. Upon receipt of a Safety Injection signal, the CCW heat exchanger inlet valve EF-HV-51 will automatically OPEN to supply increased flow for the increased heat loads on the CCW system due to the ECCS actuation.

D. Incorrect. Plausible because EF HV-59 automatically repositions on a Safety Injection signal.

However the valve CLOSES upon receipt of the SI Signal in order to maximize flow to the containment coolers.

Technical Reference(s): Drawing J-22EF05, Essential Service Water System Motor Operated Isolation Valves. REV 2 References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-05, Obj. D. Describe the operation of the ESW system under the following conditions: 2. Safety Injection Signal.

Question Source: Bank # ______

Modified Bank # ___L4404___

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

K/A Match: Candidate must apply knowledge of SIS setpoints to determine that a Safety Injection has occurred, and in response to a Safety Injection monitor the correct response of the ESW system to supply flow to the appropriate emergency heat loads, the CCW heat exchanger is one of the emergency heat loads.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 076 Service Water Group # 1 K/A # K2.01 Importance Rating 2.7 Knowledge of bus power supplies to the following: Service water.

Question #53 Given the following plant conditions:

• The Plant is at 50% power

• Circwater Pumps A & C are running

• Service Water Pumps A & B are running

• A Lockout occurs on Transformer XPB122 Power to the B Service water pump will be restored when PB122 is

____(1)____ aligned to receive power from ____(2)____.

A. (1) manually (2) PB121 B. (1) manually (2) PB123 C. (1) automatically (2) PB121 D. (1) automatically (2) PB123 Answer: C Explanation:

A. Incorrect. For the conditions given, PB122 will automatically cross tie to PB121.

B. Incorrect. PB122 will only receive power from PB121 on the cross tie, however will supply power to PB123 automatically if the same conditions mentioned in explanation C are met for PB123.

C. Correct. With no bus fault on PB122, and the feeder breakers for both circwater pump and service water pump open, PB122 will automatically cross tie only to PB121 in the event of a 13.8kv feeder breaker trip or Transformer lockout.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator D. Incorrect. PB122 will only automatically cross tie to PB123 if the fault occurs on XPB123 and power is being supplied from PB122, otherwise PB122 will always tie to PB121.

Technical Reference(s): Drawing 8600-X-88550, Circ. And Servcie Water Pump House One Line Diagram 4160V, OTN-PB-00001 Addendum 3, Energizing and Cross-Tying Buses PB121, PB122, and PB123.

References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-02, Obj. G. Describe the interlocks associated with the following cross-tie breakers: PB-121, PB-122, PB-123.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 078 Instrument Air Group # 1 K/A # K4.02 Importance Rating 3.2 Knowledge of IAS design feature(s) and/or interlock(s) which provide for the following:

Cross-over to other air systems.

Question #54 Given the following plant conditions:

• The plant is at 100% power.

• The BOP observes Instrument Air Header pressure is lowering.

Which of the following describes the operation of the Service Air Header pressure control valve, KAPV0011 for this event?

A. Receives a signal to CLOSE at 105 psig and lowering to isolate a potential leak in the Service Air Header B. Receives a signal to CLOSE at 110 psig and lowering to isolate a potential leak in the Service Air Header C. Receives a signal to OPEN at 105 psig and lowering to assist in restoring the Instrument Air Header pressure D. Receives a signal to OPEN at 110 psig and lowering to assist in restoring the Instrument Air Header pressure Answer: B Explanation:

A. Incorrect. The service sir system is automatically isolated at 110 psig, however the lag air dryer will go into service at 105 psig.

B. Correct. The instrument air and service air systems are normally connected. The service air system is automatically isolated at 110 psig by closure of KA PV0011, Compress Air Sys Serv Air Sply Press Strl Vlv.

C. Incorrect. KAPV0011 does not receive any automatic open signals, only auto close signals, and it isolates the service air header to assist in maintaining the instrument air header pressure.

D. Incorrect. KAPV0011 does not receive any automatic open signals, only auto close signals, and it isolates the service air header to assist in maintaining the instrument air header pressure.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): OTN-KA-00001, Compressed Air System, Rev 26 References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-14 Obj. E. Describe the actions that occur as air pressure falls from 120 to 100 psig.

Question Source: Bank # ___L16394___

Modified Bank # ______

New _______

Question History: Last NRC Exam ______2007 Callaway ILT NRC Exam______

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 103 Containment Group # 1 K/A # A4.01 Importance Rating 3.2 Ability to manually operate and/or monitor in the control room: Flow control, pressure control, and temperature control valves, including pneumatic valve controller.

Question #55 Given the following plant conditions:

• The plant is in Mode 4 for a refueling outage.

• The crew is preparing to place Containment Mini-purge in service.

• At this time NEITHER Aux Building Normal Exhaust fan is running.

What action is required to ensure that backflow from containment to the Aux Building does NOT occur during the containment purge?

A. Ensure GK HIS-47, ACCS CTRL EXHAUST FAN 2A is secured.

B. Ensure GL HIS-32, FUEL/AUX BLD NORM EXHAUST DAMPER, is closed.

C. Ensure the Mini Purge system is NOT operated in conjunction with the Shutdown Purge system.

D. Ensure the containment purge supply fan is stopped until at least ONE Aux Building Normal Exhaust fan is running.

Answer: B Explanation:

A. Incorrect. - Plausible because having this fan running or the damper closed prevents the spread of contamination to other buildings (ie. the control building) The question is specifically asking to prevent the spread of contamination to the Aux Building.

B. Correct - With neither Aux Building Normal Exhaust fan, CGL03A nor CGL03B, running one of the dampers must be closed to prevent a flow path that would allow the spread of contamination to other parts of the building C. Incorrect - The Mini Purge system is not allowed to be run when the Shutdown purge system is in operation, however the reason is to prevent damage to mini-purge ductwork due to different capacities.

D. Incorrect - Containment purge can be run without an Aux Building Normal Exhaust fan running IF either GL HIS-32, damper for CGL03A, or GL HIS-33, damper for CGL03B, is closed

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): OTN-GT-00001 Section 5.1.3 References to be provided to applicants during examination: None Learning Objective: P. EXPLAIN the precautions, limitations and bases for the following components/conditions associated with OTN-GT-00001, "Containment Purge System":

5. Stopping and restarting a containment purge or vent Question Source: Bank # __X____

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2005_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

OTN-GT-00001 Step 5.1.3 ENSURE the following to prevent inadvertent backflow of contamination to other buildings:

At least one of the following - At least one of the following -

OR RUNNING: () CLOSED: ()

GL HIS-30, FUEL/AUX BLD GL HIS-32, FUEL/AUX BLD NORM EXH FAN A NORM EXH DAMPER GL HIS-31, FUEL/AUX BLD GL HIS-33, FUEL/AUX BLD NORM EXH FAN B NORM EXH DAMPER

NRC Site-Specific Written Examination Callaway Plant Reactor Operator

NRC Site-Specific Written Examination Callaway Plant Reactor Operator UNIT VENT AUX BLDG HVAC HIS CONTROL BLDG HVAC 20 TURBINE BLDG HVAC FUEL BLDG HVAC FUEL BLDG HVAC I CPIS MISC BLDG HVAC 29 28 34 20 HIS 34 CGT02 CGT01 PUMP RUNNING SIGNAL CPIS I PD-15B HEPA FILTER LOW P CHARCOAL BED HIGH 16 TEMP SHUTDOWN 200OF TSH 19 TT T-19 19 PDI PDS CHARCOAL BED HIGH PD-18 55 18 TEMP ALARM HEPA 200OF FILTER TE 19 PDIC 16 CHARCOAL FILTER PDI PDS PDS CTMT PURGE FILTER PD-15A 56 15B 15A ADSORBER UNIT FGTO1 HEPA FILTER PDI PDS PD-13 57 13 MOD EFF PREFILTER EXHAUST FROM CONTAINMENT

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 001 Control Rod Drive Group # 2 K/A # K4.14 Importance Rating 2.6 Knowledge of CRDS design feature(s) and/or interlock(s) which provide for the following: Operation parameters, including proper rod speed.

Question #56 Given the following plant conditions:

• Loop 1 Tavg channel is 573°F

• Loop 2 Tavg channel is 572°F

• Loop 3 Tavg channel is 573°F

• Loop 4 Tavg channel is 574°F

• Tref is 570°F

• Rod Control System is placed in AUTOMATIC At what speed will rods step?

A. 8 steps/min B. 32 steps/min C. 40 steps/min D. 72 steps/min Answer: C Explanation: At a difference of 3°F rods begin to step in a 8 steps/min. The rod speed is then ramped to a maximum of 72 steps/min when a difference of 5°F is reached. The circuitry uses the highest Tavg channel NOT the average.

A. Incorrect - This is the speed for a 3° difference. If the candidate incorrectly uses the average Tavg this is the answer they would compute.

B. Incorrect -32 steps per minutes is half of the difference between 8 steps and 72 steps ( i.e.

72-8 = 64 / 2 = 32) If the candidate did not add 8 to the 32 steps/min this is the answer they would compute.

C. Correct - 4° difference would yield 32 steps/min plus the 8 steps/min yield 40 steps/min D. Incorrect - this is the maximum speed. If the candidate incorrectly calculates 4° difference as the maximum instead of 5°difference this is the answer they would compute.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s): Rod control tech manual (see chart below)

References to be provided to applicants during examination: none Learning Objective: Discuss the operation of the reactor control unit, describing the following:

Signals used by the unit to determine when control rod motion is required Relation between rod speed and effective temperature error Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator 72 40 Deadband ROD SPEED 8

0 8

Lock up 40 72

-5 -4 -3 -2 -1 0 1 2 3 4 5 TREF - TAVE = Temperature Error SPEED-DIRECTION PROGRAM

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 017 In-core Temperature Monitor Group # 2 K/A # K1.02 Importance Rating 3.3 Knowledge of the physical connections and/or cause effect relationships between the ITM system and the following systems: RCS.

Question #57 The Core Thermocouples are connected to the vessel through ____(1)_____ and provide indication up to ____(2)_____.

(1) (2)

A. The reactor head instrument port 700°F B. The reactor head instrument port 2300°F C. Conduits extending from the 700°F bottom of the reactor vessel D. Conduits extending from the 2300°F bottom of the reactor vessel Answer: B Explanation: The thermocouples are enclosed in removable stainless steel sheaths. These sheaths are routed in guide tubes which position the thermocouple at the exit of the fuel assemblies. The guide tubes extend from the selected core location to the reactor vessel head seal assemblies. The individual thermocouple guide tubes are enclosed in a thermocouple port column which protrudes through the reactor head instrument port.

The thermocouples will indicate accurately from 0°F to 700°F. During accident conditions, the thermocouples will operate satisfactorily to 2300°F.

The movable detectors are driven into the reactor core through conduits extending from the bottom of the reactor vessel to a thimble seal table. The conduit forms an extension of the primary system pressure boundary.

A. Incorrect - This is the correct location, however the thermocouples will provide ACCURATE indication up to 700F, but will continue to indicate up to 2300F B. Correct - This is the correct location. The thermocouples will indicate up to 2300°F.

C. Incorrect - This is where the movable incore detectors are connected to the reactor vessel.

This is the correct temperature.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator D. Incorrect - Incorrect location and temperature.

Technical Reference(s): Thermocouple Core Cooling Monitor Technical Manual, ETP-BB-03138 DISASSEMBLY OF THE CORE EXIT THERMOCOUPLE NOZZLE ASSEMBLY (CETNA),

References to be provided to applicants during examination: None Learning Objective: T61.0110.6 Systems LP #29 INCORE INSTRUMENTATION - SR/BB Obj F EXPLAIN the basic operation of an incore thermocouple.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.3)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 028 Hydrogen Recombiner and Purge Group # 2 Control K/A # K5.01 Importance Rating 3.4 Knowledge of the operational implications of the following concepts as they apply to the HRPS: Explosive hydrogen concentration.

Question #58 Given the following plant conditions:

• The plant experienced a LOCA three hours ago.

• You have been directed to place an Electric Thermal Hydrogen Recombiner (ETHR) in service.

Which of the following is the LOWEST Containment Hydrogen volumetric concentration that would PREVENT the ETHR from being placed in service?

A. 3%

B. 4%

C. 5%

D. 6%

Answer: B Explanation: In accordance with OTN-GS-00001, step 5.5.1 IF containment hydrogen volumetric concentration is greater than or equal to 4%, do NOT place an ETHR in service. Therefore, 4% is correct as it is the LOWEST that would prevent the ETHR from being placed in service per the procedure.

A. Incorrect B. Correct, C. Incorrect D. Incorrect Technical Reference(s): OTN-GS-00001, Rev 16

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-40 Q. EXPLAIN the precautions, limitations and bases for the following processes/conditions associated with OTN-GS-00001, "Containment Hydrogen Control System":

01. Hydrogen concentration with Hydrogen Recombiner in service or being placed in service.

Question Source: Bank # ______

Modified Bank # ______

New __X_____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.5 )

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 029 Containment Purge Group # 2 K/A # A1.02 Importance Rating 3.4 Ability to predict and/or monitor changes in parameters to prevent exceeding design limits) associated with operating the Containment Purge System controls including:

Radiation levels.

Question #59 Which of the following describes the MINIMUM detector response that will generate a Containment Purge Isolation Signal (CPIS)?

A. EITHER GT RE-22 or GT RE-33, Containment Purge Exhaust Monitors, reach the HI-HI setpoint.

B. BOTH GT RE-22 AND GT RE-33, Containment Purge Exhaust Monitors, reach the HI-HI setpoint.

C. EITHER GT RE-31 or GT RE-32, Containment Atmosphere Radiation Monitors, reach the HI-HI setpoint.

D. BOTH GT RE-31 AND GT RE-32, Containment Atmosphere Radiation Monitors, reach the HI-HI setpoint.

Answer: A Explanation:

A. Correct. A High High signal on Either GT RE-22 or GT RE-33 will actuate a CPIS.

B. Incorrect. Though either GT RE-22 or GT RE-33 will actuate a CPIS, the question is asking for the MINIMUM detector response, therefore the requirement of BOTH makes this an incorrect answer.

C. Incorrect. GT RE-31 and GT RE-32 are Containment Atmospheric Radiation monitors that provide alarms and indications but do not cause any automatic actuations.

D. Incorrect. GT RE-31 and GT RE-32 are Containment Atmospheric Radiation monitors that provide alarms and indications but do not cause any automatic actuations.

Technical Reference(s): Logic Block Diagram ESFAS J-104-00176 Rev 13.

References to be provided to applicants during examination: None.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Learning Objective: T61.0110 LP-40, Obj. N. List the signals that cause a Containment Purge Isolation sisgnal (CPIS).

Question Source: Bank # ___L16318___

Modified Bank # ______

New _______

Question History: Last NRC Exam ______2007 Callaway ILT NRC Exam______

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.5)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 033 Spent Fuel Pool Cooling Group # 2 K/A # K4.01 Importance Rating 2.9 Knowledge of design feature(s) and/or interlock(s) which provide for the following:

Maintenance of spent fuel level.

Question #60 Given the following plant conditions:

• CRDM latch and drag testing is being performed following core load.

• ECV0995, Fuel Transfer Tube Isolation Valve is open.

• A cavity seal ring leak occurs.

• Refueling pool level reaches 364 inches on BB LI-53A and 53B before ECV0995 is closed.

• Spent Fuel Pool Level indicates -24 inches.

Which of the following pumps will have tripped?

A. Fuel Pool Skimmer pump.

B. Fuel Pool Cleanup pumps.

C. Residual Heat Removal pumps.

D. Spent Fuel Pool Cooling pumps.

Answer: D Explanation:

A. Incorrect. There is no trip on low level for the Fuel pool skimmer pump.

B. Incorrect. There is no trip on low level for the Fuel pool cleanup pump. This pump does have a low flow alarm, however it has no automatic trips C. Incorrect. RHR pumps do not take a suction from the spent fuel pool. This is a plausible distractor due to the actions taken during LOCA. Also Plausible because when the Refuel Pool is connected to the SFP then RHR pump cavitation could occur if level gets too low. The operators are directed to isolate the SFP heat exchanger to maximize CCW flow to the RHR heat exchangers. If the operator does not understand the reason for isolating SFP heat exchangers and securing the SFP pumps, this distractor becomes plausible.

D. Correct. The Spent fuel pool cooling pumps trip on low level in the fuel pool. The low level is at -22 inches. The condition above indicates -24 inches. At this point the spent fuel cooling pump

NRC Site-Specific Written Examination Callaway Plant Reactor Operator should have tripped.

Technical Reference(s):

1. OTN EC 00001, Fuel Pool Cooling and Clean up System Normal Operating Procedure, Rev 39

2. OOA-BB-00003 References to be provided to applicants during examination: None Learning Objective: T61.0110.6 Systems LP #24 Objective E - EXPLAIN the design features that prevent draining the Spent Fuel Pool.

Question Source: Bank # __X__R12185__

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 035 Steam Generator Group # 2 K/A # K6.03 Importance Rating 2.6 Knowledge of the effect of a loss or malfunction on the following will have on the S/GS:

S/G level detector.

Question #61 With the plant stable at 100% power, assuming no action by the crew, which of the following will cause actual Steam Generator Level to RISE?

A. S/G program level setpoint signal fails LOW B. One of the controlling S/G LEVEL transmitters fails LOW C. One of the controlling Steam FLOW transmitters fails LOW D. One of the controlling S/G PRESSURE transmitters fails LOW Answer: B Explanation:

A. Incorrect. Plausible because multiple signals in the steam generator level control system upon failing low will cause an error such that the system will respond to raise level, (level, feed flow, pressure) However if the setpoint fails low, the steam generator will see actual level High and then try to lower the level by reducing feed flow.

B. Correct. The two controlling channels for S/G level are averaged and the output is fed into the S/G level control circuitry. When one channel fails low, the average is lowered resulting in a negative error which causes the system to respond by trying to raise level to the program setpoint.

C. Incorrect. Plausible because if a Steam Pressure channel fails low, this would give a false HI Mass Flow Rate, which would result in the system attempting to make up for the increased loss of inventory in the steam generator and raising level. However if the steam flow transmitter fails low, it would have the opposite effect. To try to match steam flow (now low) and feed flow the steam generator will reduce feed flow which will lower level. With the digital feedwater system, three channels are used to determine steam flow, with one failing low, the system will then reject the failed signal and average the remaining two signals, which will result in no change in S/G water level.

D. Incorrect. Plausible because if a Steam Pressure channel fails low, this would give a false HI Mass Flow Rate, which would result in the system attempting to make up for the increased loss of inventory in the steam generator and raising level.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical Reference(s):

1. Drawing 7250D64 Sheet 13, Functional Diagram Feedwater control & isolation. REV 15

2. OTO-AE-00002, SG Water Level Control Malfunctions, Rev 13 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP #40 - SG Water Level control Malfunctions - Objective B DESCRIBE symptoms or entry conditions for OTO-AE-00002, Steam Generator Water Level Control Malfunctions.

Question Source: Bank # ___L16317___

Modified Bank # ______

New _______

Question History: Last NRC Exam ____Callaway 2007 Initial NRC Exam________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

K/A match: The question is written in an operationally valid manner such that the candidate is given a changing indication and asked to analyze each failure of a Steam Generator instrument and determine the resultant effect it would have on the Steam Generator. Knowledge of the Steam Generator Water level system and the response to failures is required to answer this question.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 041 Steam Dump/Turbine Bypass Group # 2 Control K/A # K3.01 Importance Rating 3.2 Knowledge of the effect that a loss or malfunction of the SDS will have on the following:

S/G.

Question #62 Given the following plant conditions:

• The plant was operating at 100% power when a spurious reactor trip occurs.

• The C-9 interlock fails such that it DEENERGIZES the Condenser Steam Dump Arming solenoid.

(1) Which of the following systems will be used to control RCS temperature, And (2) At what pressure will the Steam Generators be maintained?

A. (1) Condenser Steam Dumps (2) 1092 psig B. (1) Steam Generator Atmospheric Steam Dumps (2) 1092 psig C. (1) Condenser Steam Dumps (2) 1125 psig D. (1) Steam Generator Atmospheric Steam Dumps (2) 1125 psig Answer: D Explanation:

A. Incorrect. Plausible if the candidate believes the solenoid is normally energized and is deenergized to actuate and ARM the steam dump system, making 1092 psig or NO LOAD pressure plausible as this is the pressure that the condenser steam dumps will control at for a NO

NRC Site-Specific Written Examination Callaway Plant Reactor Operator LOAD Tavg. It is incorrect because the ARMING solenoid must be energized to ARM the steam dumps, therefore the condenser steam dumps are not available with this failure.

B. Incorrect. Plausible because the condenser steam dumps are not available therefore the S/G ASDs will control RCS Tavg and S/G Pressure. 1092 psig is the NO LOAD setpoint for the S/Gs and corresponds to an RCS Tavg of 557F. Under this situation the RCS will heat up due to no steam load and excessive heat following a reactor trip until the S/G ASD setpoint of 1125 psig is reached and the ASDs will open controlling pressure.

C. Incorrect. Plausible if the candidate believes the solenoid is normally energized and is deenergized to actuate and ARM the steam dump system, 1125 psig is correct for controlling on the ASDs, If the candidate incorrectly assumes the pressure will be controlled on the Load Reject controller, as it would if the P-4 failure of the Steam Dumps system had occurred, and does not convert psia to psig from the steam tables, then 1125 can be obtained incorrectly.

D. Correct. With condenser steam dumps unavailable, the S/G pressure and subsequently RCS Tavg, will be controlled with the S/G ASDs at their normal setpoint of 1125 psig.

Technical Reference(s): OTN-AB-00001, Main Steam and Steam Dump Systems, REV 22.

ASME Steam Tables, Copyright 2006 References to be provided to applicants during examination: ASME Steam Tables.

Learning Objective: T61.0110 LP-20, Obj. I. Discuss the four Steam Dump permissive interlocks and Explain the effects of each on system operation. Obj. B Describe the purpose and operation of the following Main Steam System components: 4. Steam Dumps and Isolation Valves.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

K/A Match: Candidate must understand the operation of the steam dump system arming solenoid and the expected position upon deenergizing, and then predict the impact on the Steam Generator pressure due to this malfunction.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 072 Area Radiation Monitoring Group # 2 K/A # A4.03 Importance Rating 3.1 Ability to manually operate and/or monitor in the control room: Check source for operability demonstration.

Question #63 On Area Radiation Monitor panel SD055A, which of the following describes the method for performing a source check?

A. Pushing the Source Check button on the key pad B. Pushing the red indicating light push button on the ARM module C. Pushing the amber indicating light push button on the ARM module D. Pushing the green normal indicating light push button on the ARM module

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Answer: D Explanation:

A. Incorrect. Plausible because the RM-23 Process Rad Monitors have a key pad with a C/S pushbutton to perform a source check of the Process Rad Monitor. Incorrect because the Area Radiation Monitors do not have key pad controls.

B. Incorrect. Plausible because each color indicating light is also a push button that serves a specific purpose. Incorrect because the red indicating light will reset the HIGH alarm and turn off the local flashing light when rad levels decrease below the setpoint.

C. Incorrect. Plausible because each color indicating light is also a push button that serves a specific purpose. Incorrect because the amber indicating light will reset the ALERT alarm when rad levels decrease below the setpoint.

D. Correct. To ensure the operability of each detector channel, each detector assembly contains its own SR/Y 90 check source. This source is positioned near the detector by energizing the check source solenoid on the monitor in the Control Room. The check source is performed by depressing the normal light on the monitor and watching for system response.

Technical Reference(s): This action is considered skill of the craft based on the display and indications given to the operator. See diagram below.

SNUPPS RMS II 10,000 8

6 4

2 1000 8

6 4

2 mR/hr 100 8

6 4

2 10 8

6 4

2 1

8 6

4 2

.1 PUSH TO ALARMS CHECK RESET RESET NORMAL ALERT HIGH eberline

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination: None.

Learning Objective: T61.0110 System LP-36, Obj. C. Identify the Process and Area Radiation Monitoring Control Room controls, alarms, and indications and Describe how each is used to predict, monitor and control the Process and Area Radiation Monitoring System.

Question Source: Bank # ___L5482___

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.5)

Comments:

K/A Match: Candidate must know how to manually operate the Area Radiation Monitor controls to perform a source check to demonstrate operability.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 075 Circulating Water Group # 2 K/A # K2.03 Importance Rating 2.6 Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of Circulating water pumps.

Question #64 Given the following initial plant conditions:

• The Plant is in Mode 1 with Reactor Power at 96% power

• HP Condenser Backpressure is 4.3" HgA and stable A transient occurs and the following conditions are observed:

• Annunciator 13A, CIRC WTR PMP LOCKOUT, in alarm

• Reactor Power lowering

• Generator Load lowering

• Control Bank 'D' inserting in AUTO (1) Which of the following plant transients is in progress, And (2) Which action will be taken by the Reactor Operator?

A. (1) Turbine Setback (2) Place Rod control in Manual B. (1) Turbine Setback (2) Verify Tavg is trending to or within 5F of Tref C. (1) Rod Control Malfunction (2) Place Rod control in Manual D. (1) Rod Control Malfunction (2) Verify Tavg is trending to or within 5F of Tref Answer: B

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Explanation:

A. Incorrect. Plausible because Turbine Setback is correct, however rods should NOT be placed in manual.

B. Correct. When the HP condenser vacuum is being maintained greater than 4 HgA, the condenser Circ Pump Setback is Enabled. This feature of the Circulating Water system prevents a loss of vacuum in the condenser in the event a circulating water pump trips during operation by automatically reducing turbine load to approximately 75% power. In response to this event the crew should enter OTO-MA-00001, Turbine Load Rejection, and verify rods are responding to the Tavg/Tref mismatch caused by the load rejection.

C. Incorrect. Plausible because a Rod Control Malfunction will provide many of the same indications accompanied by an NI failure or Turbine impulse pressure alarm. The correct action for a rod control malfunction per OTO-SF-00001, Rod Control Malfunctions, is to place rods in manual.

D. Incorrect. Plausible because a Rod Control Malfunction will provide many of the same indications accompanied by an NI failure or Turbine impulse pressure alarm.

Technical Reference(s): OTN-DA-00001, Circulating Water System, REV 33, OTO-MA-00001, Turbine Load Rejection, REV 29 References to be provided to applicants during examination: None.

Learning Objective:

Question Source: Bank # ______

Modified Bank # __R15008____

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

K/A Match: Candidate must predict the impact of a Circulating Water Pump trip on the Turbine Setback feature of the Circwater System while it is enabled, and identify the action required to mitigate the consequences of this event.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 086 Fire Protection Group # 2 K/A # A3.03 Importance Rating 2.9 Ability to monitor automatic operation of the Fire Protection System including: Actuation of fire detectors.

Question #65 Given the following plant conditions:

• The plant is in Mode 4

• KC-008 indicates Halon actuated in Electrical Pen Room A Which of the following combinations of signals would have ACTUATED the Halon system?

A. Detector 3 in alarm, and detector 13 in alarm B. Detector 5 in alarm, and detector 2 has a trouble alarm C. Detector 10 has a trouble alarm, and detector 13 in alarm D. Detector 4 has a trouble alarm, and detector 15 has a trouble alarm

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Answer: A Explanation:

A. Correct. A cross-zoned detector circuit. Sixteen detectors in a 4 by 4 array are connected so that adjacent detectors are on different circuits, or "zones". In order for the Halon system to automatically actuate, detectors in both loops must sense a fire or a detector in one loop senses a fire while a trouble signal is present on the other loop. Detection of a fire by one loop without a detection or trouble signal in the other loop will give an alarm only.

B. Incorrect. See explanation A.

C. Incorrect. See explanation A.

D. Incorrect. See explanation A.

Technical Reference(s): OTA-KC-01008 Addendum 6, KC-008 Message File REV 11 References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-35, Obj. E. List the automatic system operation signals including pump starting, spray actuation, trouble alarms and fire alarms.

Question Source: Bank # ___L16658___

Modified Bank # ______

New _______

Question History: Last NRC Exam _____2009 Callaway ILT NRC Exam_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.7)

Comments:

Hi Cog question because candidate must know logic for actuation and apply it to the map to determine if logic is met in each of the answers.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Conduct of Operations Group # Generic K/A # 2.1.5 Importance Rating 2.9 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Question #66 Given the following plant conditions:

• The plant is in Mode 1.

• The shift is staffed ONLY to the MINIMUM required.

• The shift has 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> remaining.

• The RO has become ill and must leave the site for emergency medical treatment.

Which of the following describes the requirements regarding the shift composition and the MINIMUM required action in this situation?

A. The RO may NOT leave until another qualified RO is on site.

B. Responsibilities of the RO may be turned over to the BOP for the remainder of the shift.

C. The RO may leave the site immediately and a replacement must arrive within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. The RO may leave the site immediately and a replacement must arrive within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

Answer: D Explanation:

A. Incorrect. Plausible if the candidate assumes that minimum staffing must always be maintained. Incorrect because minimum staffing can be below the minimum for no more than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

B. Incorrect. This would place the shift crew below the minimum staffing requirements for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> which exceeds the maximum of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> below minimum staffing required by Tech Specs.

Plausible if the candidate assumes the maximum time is 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Incorrect. At minimum staffing for the shift crew, this would still place the crew below the minimum required staffing. Plausible if the candidate assumes that minimum staffing must be restored within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. Correct. T/S 5.2.2 states that the Shift crew composition may be one less than the minimum requirement for a period of time not to exceed 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements.

Technical Reference(s): ODP-ZZ-00001, Operations Department Code of Conduct, Rev 91 References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-66, Obj. A.7. Explain the following as applied in ODP-ZZ-00001, Operations Department Code of Conduct; Discuss Minimum Shift Manning Requirements, Unexpected absence requirements regarding shift complement.

Question Source: Bank # __L16465____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____2007 Callaway ILT NRC Exam________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.1)

Comments:

K/A match: The question tests the candidates ability to apply the requirements for minimum staffing as directed in ODP-ZZ-00001 and Technical Specifications Section 5.2.2 to a realistic situation that could occur while on shift.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Conduct of Operations Group # Generic K/A # 2.1.26 Importance Rating 3.4 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

Question #67 Given the following plant conditions:

• A permit required confined space entry is to be conducted at the Water Treatment Plant blowdown line manhole.

In accordance with APA-ZZ-00802, Confined Space Entry Permit Program; (1) Who must be present at the confined space entry while Entrants are in the space?

And (2) Who will enter the space if an emergency rescue must be performed?

A. (1) Attendant (2) Fire Brigade B. (1) Attendant (2) Medical Emergency Response Team C. (1) Entry Supervisor (2) Fire Brigade D. (1) Entry Supervisor (2) Medical Emergency Response Team Answer: A Explanation:

A. Correct. IAW APA-ZZ-00802, when entry into a Permit Required Confined Space is performed, the Attendant must always be present when entrants are in the space, and the Fire Brigade will serve as the Rescue Team if a non-entry retrieval can NOT be performed.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Incorrect. Plausible because the Attendant must be present, and the MERT is briefed of the entry prior to entry, however the Fire Brigade serves as the Rescue Team for emergency retrieval.

C. Incorrect. Plausible because the Entry Supervisor has the responsibility to ensure all individuals are qualified and briefed on the confined space entry, however he is not required to remain present during the entry. The Fire Brigade will serve as the Rescue Team and enter if a non-entry retrieval can NOT be performed.

D. Incorrect. Plausible because the Entry Supervisor has the responsibility to ensure all individuals are qualified and briefed on the confined space entry, however he is not required to remain present during the entry. The MERT is briefed of the entry prior to entry, however the Fire Brigade serves as the Rescue Team for emergency retrieval.

Technical Reference(s): APA-ZZ-00802, Confined Space Program, REV 18 References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems LP-72, Obj. F.4. In accordance with APA-ZZ-00802, Confined Space Program, Describe the requirements to enter a Permit Required Confined Space.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.10)

Comments:

K/A match: Knowledge of the Industrial Safety procedure for confined space entry required to answer the question. The confined space entry program addresses issues with oxygen levels, and toxic gasses which can be found in certain spaces.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Conduct of Operations Group # Generic K/A # 2.1.34 Importance Rating 2.7 Knowledge of primary and secondary plant chemistry limits.

Question #68 What is the Technical Specification Secondary Specific Activity limit for Dose Equivalent I-131?

A. < 0.1 µCi/gm B. < 1 µCi/gm C. < 60 µCi/gm D. < 225 µCi/gm Answer: A Explanation:

A. Correct. T/S 3.7.18 states the specific activity of the secondary coolant shall be < 0.10 µCi/gm Dose Equivalent I-131.

B. Incorrect. Plausible because this is the limit for the T/S RCS Dose Equivalent I-131 limit.

C. Incorrect. Plausible because this is a T/S action limit value for RCS Dose Equivalent I-131.

D. Incorrect. Plausible because this is the T/S Dose Equivalent Xe-133 limit for the RCS.

Technical Reference(s): Technical Specifications 3.7.18, Secondary Specific Activity References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems, LP-20, Obj. K. State the Tech. Spec/FSAR LCOs associated with the Main Steam System Technical Specifications.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.10)

Comments:

K/A match: Knowledge of Secondary Plant chemistry limits (including activity) are required to correctly answer this question.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Equipment Control Group # Generic K/A # 2.2.39 Importance Rating 3.9 Knowledge of less than or equal to one hour Technical Specification action statements for systems.

Question #69 Given the following plant conditions:

• Callaway plant is coasting down at the end of core life.

• Reactor power is currently 82%.

• A transient occurs resulting in Axial Flux Difference (AFD) on two excore channels indicating +15%.

Which one of the following describes the AFD status and required action?

Reference Provided A. AFD is within the limit. Restore AFD to within 1% of the target value to remove the existing OTDT setpoint penalty.

B. AFD is within the limit. Restore AFD to within 1% of the target value within 30 minutes to preclude an axial xenon oscillation.

C. AFD is outside the limit. Reduce Reactor power by 3% for every 1% AFD is outside the limit within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

D. AFD is outside the limit. Restore AFD to within the limit or reduce power to

< 50% in the next 30 minutes.

Answer: D A. Incorrect - Plausible if the candidate looks at -15%, then AFD is within the limit at this power level, However at +15% AFD is outside the limit.

B. Incorrect - Plausible if the candidate looks at -15%, then AFD is within the limit at this power level, However at +15% AFD is outside the limit. There is no 30 minute requirement to restore AFD to preclude a xenon oscillation.

C. Incorrect - Plausible because AFD is outside the limit, however the requirement to reduce power by 3% for every 1% outside limit is the T/S requirement if QPTR is outside the limit.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator D. Correct - AFD is outside the limits for this power level. The correct action is to Restore AFD to within limits or reduce reactor power to less than 50% within 30 minutes.

Technical Reference(s): TS 3.2.3, Axial Flux Difference (AFD)

References to be provided to applicants during examination: Curve Book Figure 1-1 for AFD.

Learning Objective: T61.0110 LP-67, Obj. 2. State the Limiting Conditions for Operation and Explain the Applicability, actions, and bases for the following: C. T/S 3.2.3 AFD.

Question Source: Bank # ___R8597___

Modified Bank # ______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

CFR: 41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Equipment Control Group # Generic K/A # 2.2.13 Importance Rating 4.1 Knowledge of tagging and clearance procedures.

Question #70 An Operations Technician is in the process of performing an Independent Verification (IV) of Workmans Protection Assurance (WPA) placed in the plant.

Which of the following radiation levels would permit the Independent Verification (IV) to be waived?

A. Airborne activity in the area is 1.5 DAC B. General Area Dose Rate is 15 mRem/Hr C. Contamination in the area is 15 dpm/100 cm2 D. An Exposure of 15 mRem is likely to be received Answer: D Explanation: Per the technical reference, the waiver of an IV is allowed when radiation exposures is expected to be greater than 10mR.

A. Incorrect. There is no specific limit on airborne activity.

B. Incorrect. The value for general dose rates is 25mR/hr.

C. Incorrect. There is no specific limit on area contamination.

D. Correct Technical Reference(s): ODP-ZZ-00310 Rev 60. Note on page 30 that states The waiver of an IV is allowed when General Area Dose Rates are greater than 25mRem/Hr, or in situations where radiation exposures of greater than 10 mRem are likely.

References to be provided to applicants during examination: None Learning Objective: T61.003A, LP #13, WPA.

A. ODP-ZZ-00310, WPA and Caution Tagging

1. DISCUSS the Responsibilities of the following:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator

a. Shift Manager/Control Room Supervisor

1. Discus when an Independent Verification may be waived.

Question Source: Bank # __X R6464____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.10)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Equipment Control Group # Generic K/A # 2.2.44 Importance Rating 4.2 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.

Question #71 Given the following plant conditions and sequence of events:

• A Reactor Startup is in progress.

• Reactor Power is 1x10-8 amps.

• D Bank is at 105 steps.

• The RO is directed to raise power to 4%.

• The RO establishes a positive startup rate and releases the IN-HOLD-OUT switch.

• Rods continue to step outward as indicated on group step counters and DRPI.

The immediate required action is to __ (1) _.

And If NO operator action is taken, the consequence of inaction is that reactor power will rise to _ (2) _before being automatically terminated by the Reactor Protection System.

(1) (2)

A. place Control Rod Group Selector 10%

to Bank D position and insert control rods to zero steps.

B. place Control Rod Group Selector 25%

to Bank D position and insert control rods to zero steps.

C. manually trip the reactor 10%

D. manually trip the reactor 25%

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Answer: D Explanation:Per OTO-SF-00001 Rod Control Malfuctions the immediate action would be to place Rod Control in MANUAL. Based on the information given the Rod Control is already in MANUAL. The next step is to Manually TRIP the Reactor. If this action is not take reactor power will continue to increase until the Reactor is tripped at the low power trip setpoint of 25%.

A. Incorrect - incorrect action and power level for the reactor trip. 10% is the P-10 permissive which will enable reactor trips on Pressurizer and RCP parameters, but these trips will not be met prior to reaching the 25% Intermediate Range or Low Power Range trip.

B. Incorrect - incorrect action. Correct power level for the RPS reactor trip.

C. Incorrect - correct action, incorrect power level for the reactor trip. 10% is the P-10 permissive which will enable reactor trips on Pressurizer and RCP parameters, but these trips will not be met prior to reaching the 25% Intermediate Range or Low Power Range trip.

D. Correct - correct action and correct power level Technical Reference(s): OTO-SF-00001 References to be provided to applicants during examination: None Learning Objective: LIST and DESCRIBE all reactor trip signals.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____X_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 41.10 / 45.12 / 45.13)

Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Radiation Control Group # Generic K/A # 2.3.4 Importance Rating 3.2 Knowledge of radiation exposure limits under normal or emergency conditions.

Question #72 During a plant emergency, which of the following positions is the LOWEST level of authority that can authorize dose exposure in excess of the limits of 10CFR20.1201, Occupational Dose Limits for Adults?

A. Emergency Coordinator B. Director, Nuclear Operations C. Manager, Radiation Protection D. Senior Director, Nuclear Operations Answer: A Explanation:

A. Correct. Only the Senior Vice President/CNO, Vice President of Nuclear Operations, Emergency coordinator, or Recovery Manager have the authority to authorize exceeding the federal dose limits of 10CFR20.1201 under specific conditions delineated in APA-ZZ-01000, and HDP-ZZ-01450.

B. Incorrect. Plausible because the Director, Nuclear Operations is responsible for the safe, legal, and efficient operation of the Callaway plant and oversight of all operations personnel.

C. Incorrect. Plausible because the Radiation Protection Manager has the final say for multiple radiation protection processes and procedures, however they cannot authorize an individual to exceed the federal dose limits.

D. Incorrect. Plausible because the Senior Director of Nuclear Operations may authorize the waiving of radiological requirements (ie. not maintaining respirator fit qualification current) of our site procedures for specific personnel with valid reason. However the Plant director has no authority to allow exceeding the federal dose limits.

Technical Reference(s): HDP-ZZ-01450, Authorization to Exceed Federal Occupational Dose, Rev 11.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-75, Obj. I. HDP-ZZ-01450, Authorization to Exceed Federal Occupational Dose; 1. Identify who can authorize dose exposure in excess of 10CFR20.1201 dose limits.

Question Source: Bank # __L15145____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.12)

Comments:

K/A Match: The candidate must display knowledge of the procedures used during emergency situations regarding exposure limits.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Radiation Control Group # Generic K/A # 2.3.11 Importance Rating 3.8 Ability to control radiation releases.

Question #73 Given the following plant conditions:

• The plant was at 100% power when a Steam Generator Tube Rupture occurred on D Steam Generator.

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

In accordance with E-3, to minimize radiation releases the D Steam Generator Atmospheric Steam Dump (ASD) will be _____________.

A. placed in MANUAL and CLOSED B. placed in MANUAL and controlled at 1160 psig C. adjusted in AUTO to control pressure at 1160 psig D. adjusted in AUTO to control pressure at 1185 psig Answer: C Explanation:

A. Incorrect. Plausible because this would prevent any releases due to the ASD opening, however if an actual overpressure condition occurs in the S/G this will cause a challenge to the S/G safety valves, potentially resulting in a larger release than if the ASD opened. This is also the RNO action if the ASD is open while below the selected setpoint for Automatic Control.

B. Incorrect. Plausible because 1160 psig is the desired pressure to control the S/G ASD at, however Manual control is not desired due to the attention it requires the operator place on the single parameter which could distract from the overall goal of E-3 to cool down the plant to minimize the effects of the S/G tube rupture.

C. Correct. In accordance with E-3, The ruptured steam generator is isolated to minimize radiation releases, including adjusting the steam generator ASD setpoint to 1160 psig to minimize the cycling of the ASD while keeping the lift setpoint below that of the SG Safety valves. Though it is not explicitly directed to be placed in AUTO, it is implied in that the setpoint adjustment only controls the ASD in the Automatic mode. If the ASD is not closed below the selected setpoint, then the RNO action is to place the controller in Manual and close the ASD.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator D. Incorrect. Plausible because 1185 psig is the setpoint for the first S/G Safety valve opening.

The candidate could believe that setting the ASD equal to the Safety valve will prevent inadvertent release, while still maintaining S/G integrity. Incorrect because the ASD is adjusted to 1160 psig to prevent Inadvertant release while also preventing the S/G safety from opening.

Technical Reference(s): E-3, Steam Generator Tube Rupture Rev 14 References to be provided to applicants during examination: None.

Learning Objective: T61.003D LP-D-17, Obj. I. Outline the procedural flowpath including major system and equipment operation in accomplishing the goal of E-3, Steam Generator Tube Rupture.

Question Source: Bank # ___Vogtle Unit 1 2012 SRO Exam___

Modified Bank # ______

New _______

Question History: Last NRC Exam _____Vogtle 2012 NRC SRO exam______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.11)

Comments:

K/A Match: Candidate must display knowledge of actions taken in plant procedures to minimize radiation releases to the environment.

Not currently in the Callaway Exam Bank. Question was changed to plant specific information to fit Callaway procedures from a question from the 2012 Vogtle NRC Exam.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Emergency Procedures / Plan Group # Generic K/A # 2.4.34 Importance Rating 4.2 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects.

Question #74 Given the following plant conditions:

• The plant is at 100% power.

• A fire in the Control Room substantially reduced visibility in the room.

• The Control Room is evacuated IAW OTO-ZZ-00001, Control Room Inaccessibility.

• The Crew has stabilized the plant from the Aux Shutdown Panel.

(1) Which of the following describes the AFW Pump lineup, And (2) The affect that this lineup will have on the plant?

A. (1) "A" Motor Driven AFW Pump in service; AND (2) The "B" Steam Generator Level will rise.

B. (1) "A" Motor Driven AFW Pump in service; AND (2) The "C" Steam Generator Level will rise.

C. (1) "B" Motor Driven AFW Pump in service; AND (2) The "A" Steam Generator Level will rise.

D. (1) "B" Motor Driven AFW Pump in service; AND (2) The "D" Steam Generator Level will rise.

Answer: D Explanation:

A. Incorrect. Plausible if the candidate assumes that A Train components are operated from the ASP. A MDAFW pump is locally inhibited from starting. Control at the ASP is for the B train components including the B MDAFW pump. The B S/G level will increase however it is being fed from the TDAFW pump.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Incorrect. Plausible if candidate assumes the A train components are operated from the ASP.

A MDAFW pump is locally inhibited from starting. C S/G is isolated and not used for the plant cooldown therefore level will not be rising in the C S/G.

C. Incorrect. Plausible because B MDAFW pump will be running and is capable of feeding A S/G, however only the B train components are operated from the ASP, therefore only B & D S/G ASDs can be controlled to control the plant cooldown, therefore A S/G is isolated and level will not be rising.

D. Correct. OTO-ZZ-00001, directs the operators to locally align plant equipment to support plant cooldown with a desired final lineup of B MDAFW pump supplying D S/G, and the TDAFW pump supplying the B S/G, A & C S/G ASDs are manually isolated preventing the need to feed A or C S/G.

Technical Reference(s): OTO-ZZ-00001, Control Room Inaccessibility, Rev 40 References to be provided to applicants during examination: None.

Learning Objective: T61.003B LP-B -31, Obj. E. Given a set of plant conditions or parameters indicating Control Room Inaccessibility, Identify the correct procedure to be utilized and Outline the high level actions to stabilize the plant.

Question Source: Bank # ___L16704___

Modified Bank # ______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.10)

Comments:

K/A match: Knowledge of the overall sequence of events, mitigating strategy, and operator actions for RO/BOP/Safe Shutdown Operator are required to answer this question. Knowing that the BOP will report to the Aux. Shutdown Panel and that the procedure directs the final lineup to have B MDAFW pump running and feeding D S/G is RO knowledge for this procedure.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Emergency Procedures / Plan Group # Generic K/A # 2.4.43 Importance Rating 3.2 Knowledge of emergency communications systems and techniques.

Question #75 Given the following plant conditions:

• A Site Area Emergency has been declared IAW EIP-ZZ-00101, Classification of Emergencies.

• The SENTRY notification system is NOT available.

In accordance with EIP-ZZ-00201 Addendum A, Control Room Notification Flowchart, which of the following methods will be used to notify the State and EPZ counties?

A. Plant Security Radio B. Control Room Satellite Telephone C. Backup Radio System (BURS)

D. Commercial Telephone System Answer: C Explanation:

A. Incorrect. Plausible because it is an emergency communication device available in the control room. The plant radio system is used for normal communications from the control room to operators in the field, and other security personnel.

B. Incorrect. Plausible because it is an emergency communication device available in the control room. The satellite phone is used as a last resort backup to the telephone and radio systems.

C. Correct. The BURS is an 800 MHz system used to communicate with the State and EPZ Counties when SENTRY is unavailable.

D. Incorrect. Plausible because this is a communication device that is available in the control room. The commercial telephone system is used for standard communications over the open commercial lines. EIP-ZZ-00201 directs the use of the BURS if Sentry is not available.

Technical Reference(s): EIP-ZZ-00201 Addendum A, Control Room Notification Flow Chart,

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Rev 22 References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-41, Obj. B. List the communications systems available in the control room and given specific operating conditions, Select the most appropriate control room communications System for the evolution.

Question Source: Bank # ___R16163___

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 41.10)

Comments:

K/A Match: Knowledge of each of the available plant communication systems is required to answer this question, along with knowledge of the EIP procedures used in an emergency situation.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 000007 (BW/E02&E10; CE/E02) Group # 1 Reactor Trip - Stabilization - Recovery /

1 K/A # 2.4.45 Importance Rating 4.3 Ability to prioritize and interpret the significance of each annunciator or alarm.

Question #76 Given the following plant conditions:

• A Reactor Trip and Safety Injection have occurred and the crew is performing actions contained in E-0, Reactor Trip or Safety Injection.

• RCS pressure is 1600 psig and lowering.

• Pressurizer level is 7% and rising.

• SG pressures are 900 psig and stable.

• SG levels are being controlled by AFW, with NR SG levels at 26%.

Then, the following annunciators are received:

• ANN 34D, Pressurizer Relief Tank Temperature High

• ANN 34E, Pressurizer Relief Tank Pressure High

• ANN 47C, RWST Level Low-Low 1 Auto Transfer

• ANN 62B, Area Radiation High While the Reactor Operator is reporting the above information, Annunciator 34E clears.

Based on these conditions, the Control Room Supervisor should FIRST transition to?

A. ES-1.1, SI Termination.

B. ECA-1.2, LOCA Outside Containment.

C. ES-1.3, Transfer to Cold Leg Recirculation.

D. E-1, Loss of Reactor or Secondary Coolant.

Answer: C

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Explanation: The foldout page criteria for swapover to cold leg recirc applies and the CRS should choose to transition to ES-1.3, Transfer to cold leg recirculation. Annunciator 47C provides the information that RWST level is less than 36% level and the swapover is required.

This is a priority for the control room and takes precedence per ODP-ZZ-00025 step 4.20.5.

which states The Foldout Page contains several pieces of information or actions which are normally applicable at any step of the particular procedure. The most important of these actions are procedure transitions which allow immediate response to new symptoms as they appear. The placement of these transitions on the Foldout Page allows prompt response to the appearance of subsequent symptoms.

Per Step 17 of E-0, PZR level is to low to transition to ES - 1.1, 9% PZR level would be required.

Step 21 of E-0 directs the operator to check PRT conditions Normal, the RNO says to evaluate the cause of the abnormal conditions. The candidate may choose this as the above indications indicate a LOCA through the PZR to the PRT to containment.

The annunciator for Area Rad High indicates that there may be a LOCA outside containment per step 20 RNO of E-0.

A. Incorrect - Foldout page action takes priority, and PZR level is to low for SI termination B. Incorrect - Foldout page action takes priority, C. Correct - See above D. Incorrect - Foldout page action takes priority, and while actions in E-0 step 11 may lead you to E-1 due to stuck open PZR PORV the foldout page takes priority. Also the PRT indication may result in a candidate taking this path due the E-0 step #21 directs only to evaluate the cause of the abnormal indications Technical Reference(s):

1. E-0, Reactor Trip or Safety Injection, Rev 16

2. BD-E-0, Basis Document for E-0, Rev 6

3. Annunciator 34D, Pressurizer Relief Tank Temperature High, Revision 0

4. Annunciator 34E, Pressurizer Relief Tank Pressure High, Revision 0

5. Annunciator 47C, RWST Level Low-Low 1 Auto Transfer, Revision 2

6. Annunciator 62B, Area Radiation High, Rev 1

7. ODP-ZZ-00025, Rev 25, EOP / OTO Users Guide References to be provided to applicants during examination: None Learning Objective:

1. T61.003D Emergency Operations - LP #D-4, E-0 Objective K - STATE and EXPLAIN the parameters which are evaluated, including their Criteria and Basis, to transition from E-0 to other procedures.

2. T61.003D Emergency Operations - LP #D-4, E-0 Objective L - OUTLINE procedural flowpath including major system and equipment operation in accomplishing the goal of E-0.

Question Source: Bank # _________

Modified Bank # _X __ L16726___

New _______

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

10 CFR 55.43.5 - Assessment of facility conditions and selection of the appropriate procedures during normal, abnormal, and emergency conditions.

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 000015/17 RCP Malfunctions / 4 Group # 1 K/A # 2.1.7 Importance Rating 4.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Question #77 The plant is operating at 40% power.

The A Reactor Coolant Pump (RCP) must be secured due to HIGH vibration per OTO-BB-00002, Reactor Coolant Pump Off-Normal.

What action is required to be taken to mitigate the consequences of this abnormal RCP shutdown?

A. Close the #1 Seal Leakoff for the A Reactor Coolant Pump after it has come to a stop B. Trip the Reactor, Turbine, and A RCP. Transition to E-0, Reactor Trip or Safety Injection C. Declare Pressurizer Spray Valve A inoperable. Control RCS pressure with Pressurizer PORVs D. Defeat Tavg and Delta T for the idle loop and transition to OTO-MA-00008, Rapid Load Reduction.

Answer: D Explanation: OTO-BB-00002, RCP OFF-Normal Attachment E is the applicable procedure for the conditions given.

A. Incorrect - This would be the correct action if seal leakage was greater then 6 gpm B. Incorrect - This would be the correct action if reactor power was greater than 48% when the event started.

C. Incorrect - RCS pressure would be controlled with the Loop 2 PZR spray.

D. Correct -The correct action to take is to defeat Tavg And T For Idle RCS Loop. Per TS the unit must be placed in MODE 3 with less than the required operable RCS loops in MODE 1.

Technical Reference(s):

1. OTO-BB-00002, RCP OFF-Normal Rev 31

2. Technical Specification Bases for 3.4.4

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP #11 OTO-BB-00002 RCP Off Normal Objective D Given a set of plant conditions or parameters indicating a RCP Off-Normal condition, ANALYZE the correct procedure(s) to be utilized and the required actions to stabilize the plant.

Objective F. EXPLAIN the vibration levels at which action must be taken and DESCRIBE the required actions.

Question Source: Bank # __L14597____

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

10 CFR 55.43.5 - Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

10 CFR 55.43.1 - Conditions and limitations in the facility license.

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 000056 Loss of Off-site Power / 6 Group # 1 K/A # 2.2.12 Importance Rating 4.1 2.2.12 Knowledge of surveillance procedures.

Question #78 Given the following plant conditions:

• The plant is at 100% power.

• Annunciator 134D, Switchyard Voltage High Low, alarms. MSEI0084 is reading 327 kV.

• The Transmission Supervisor reports that the Real Time Contingency Analysis Category 8 Alarm is in. Predicted voltage is 326.6 kV.

(1) Offsite Power is ___________

And (2) The safety design basis of the AC distribution system is _____(2)________?

(1) (2)

A. OPERABLE met because the safety analysis assumes a complete loss of offsite power B. INOPERABLE met because the safety analysis assumes a complete loss of offsite power C. OPERABLE met because the safety analysis assumes that the load tap changers will restore voltage to acceptable values on NB01/02 D. INOPERABLE met because the safety analysis assumes that the load tap changers will restore voltage to acceptable values on NB01/02 Answer: B Explanation: Per technical specifications bases page B 3.8.1 - 2 and 3, an assumed loss of all offsite or all onsite AC power is included in the analysis as an initial condition of DBA

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator and transient analyses. The load tap changers are not applicable as the loss of offsite power is an initial assumption. The LTC transformers ensure off-site source operability for grid conditions that would reduce voltage to 95% of nominal or 328KV and plant conditions are below this reading. Per OSP-NE-00001, step 3.3.3.a the shift manager shall declare the offsite source INOPERABLE based on the data provided (category 8 alarm).

This is a K/A match as the knowledge of offsite power operability is from this operations Surveillance Procedure (OSP). Operability determinations are a SRO function.

A. Incorrect - offsite power shall be declared INOPERABLE B. Correct - see above explanation C. Incorrect - offsite power shall be declared INOPERABLE and the load tap changers are not apart of the safety analysis and voltages are below where the load tap changers can function to restore NB01/02 volts.

D. Incorrect - the load tap changers are not apart of the safety analysis and voltages are below where the load tap changers can function to restore NB01/02 volts.

Technical Reference(s):

1) OSP-NB-00001, Class 1E Electrical Source Verification Revision 38, step 3.3.3, Attachment 4 and Attachment 12

2) OTA-RK-00026 ADD 134D, Switchyard Voltage High Low, Rev 2

3) 3.8.1 Technical Specification Bases References to be provided to applicants during examination: None Learning Objective: T61.0110.6 Systems LP #1 MD- Switchyard Objective G DESCRIBE the two offsite independent circuits which satisfy the Tech Specs offsite power requirements.

Question Source: Bank # ______

Modified Bank # ______

New ____X___

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

CFR: 43.2 SRO knowledge due to the candidate making Operability determination which is a SRO only function and SRO Level knowledge as described under1 0CFR55.43(b) 2 - Facility operating limitations in the technical specifications and their bases.

Comments:

See page 70 of lesson plan for information

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 000057 Loss of Vital AC Inst. Bus / 6 Group # 1 K/A # AA2.19 Importance Rating 4.3 Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus.

Question #79 The plant is at 100% power when the following occurs:

• Annunciator 25A, NN01 Inst Bus UV, is LIT

• Annunciator 83C, Rx Partial Trip, is LIT

• Reactor Coolant Temperature is lowering Based on the above indications, what equipment is considered OPERABLE?

A. LSELS B. CREVS C. ESW Pump PEF01A D. Emergency Diesel NE02 Answer: D Explanation: The answer is found in the various attachments of OTO-NN-00001, Loss of safety related instrument power.

Attachment A, Loss of NN01 Attachment F, Loss of NN02 Attachment K, Loss of NN03 Attachment P, Loss of NN04 A. Incorrect - There is a NOTE in Attachment A, F, K, and P, ALL 4 NN busses are required to be energized when either train of LSELS is required to be Operable. Reference Tech Spec 3.3.5, 3.8.1 and 3.8.2.

B. Incorrect - There is a NOTE in Attachment A that states Emergency Diesel NE01, ESW Pump PEF01A, Control Room AC Unit SGK04A, CREVS, and Train A UHS Cooling Tower Train are inoperable while NN01 is deenergized. CREVS would also be inoperable if NN04 was lost C. Incorrect - There is a NOTE in Attachment A that states Emergency Diesel NE01, ESW Pump PEF01A, Control Room AC Unit SGK04A, CREVS, and Train A UHS Cooling Tower Train are inoperable while NN01 is deenergized.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator D. Correct - When NN01 is deenergized, Emergency Diesel Generator NE01 is Inoperable, however if NN02 is deenergized then NE02 is Inoperable, therefore NE02 is still Operable when NN01 is deenergized.

Technical Reference(s):

1. OTO-NN-00001, Loss of safety Related Instrument Power, Rev 32, .

2. TS Bases 3.8.7 References to be provided to applicants during examination: None Learning Objective: T61.003B 6 LP B-27 Objective C - Given a set of plant conditions or parameters indicating a Loss of Safety Related Instrument Power, ANALYZE the correct procedure(s) to be utilized and the required actions to stabilize the plant.

Question Source: Bank # ______

Modified Bank # ______

New __X_____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

CFR: 43.5 - This is SRO due to knowledge of when to implement attachments and appendices AND/OR Knowledge of the BASES of TS Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 000065 Loss of Instrument Air / 8 Group # 1 K/A # 2.7 Importance Rating AA2.04 Ability to determine and interpret the following as they apply to the Loss of Instrument Air:

Typical conditions which could cause a compressor trip (ie. high temperature)

Question #80 Given the following plant conditions:

• The plant is at 100% power.

• CKA01A, A Air Compressor, is out of service for overhaul.

• CKA01B, B Air Compressor, trips due to low oil pressure.

• The In service Central Chiller, SGB01A, trips and the Standby Central Chiller, SGB01B, CANNOT be started.

(1) What is the status of CKA01C, C Air Compressor, And (2) The Control Room Supervisor (CRS) will direct the crew to ?

A. (1) CKA01C will shutdown due to high compressor outlet temperature (2) Reduce Turbine Load IAW OTO-MA-00008, Rapid Load Reduction B. (1) CKA01C will continue to run with ESW providing cooling water flow (2) Reduce Turbine Load IAW OTO-MA-00008, Rapid Load Reduction C. (1) CKA01C will shutdown due to high compressor outlet temperature (2) Restore Service Air IAW OTN-KA-00001, Compressed Air System D. (1) CKA01C will continue to run with ESW providing cooling water flow (2) Restore Service Air IAW OTN-KA-00001, Compressed Air System Answer: A Explanation:

A. Correct. C Air compressor is cooled by chilled water supplied by the Central Chillers so with no chillers running and no cooling water, the air compressor will trip on High Air Temperature. In

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator response to the loss of instrument air the crew will enter OTO-KA-00001, Loss of Instrument Air, which will direct a down power in accordance with OTO-MA-00008, Rapid Load Reduction.

B. Incorrect. Plausible because A and B Air compressors are cooled by ESW, so it could be believed that the Air compressor is not affected by the loss of the chiller units. The candidate can assume that with a loss of the B Air compressor and the unit being greater than 80% power, a down power may be required to support continued operation on C Air compressor Only.

C. Incorrect. Plausible because C Air compressor will trip on High Air Temperature due to the loss of cooling water. Restoration of the Service Air System procedure in OTN-KA-00001, Compressed Air System will direct starting an air compressor assuming all conditions are met to start one and realigning the system for normal operations. This cannot be done because there is no air compressor that meets all the conditions to start the compressor.

D. Incorrect. Plausible because the A and B Air compressors are cooled by ESW, so it could be believed that the C air compressor is not affected by the loss of the chiller units. Restoration of the Servcie Air system is correct assuming a loss of air has occurred due to the trip of the B air compressor, but can be restored if the C Air compressor is still running.

Technical Reference(s): OTO-KA-00001, Loss of Instrument Air, Rev 20 References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-14, Obj. B. Describe the purpose and operation of the following Service and instrument Air components, 1) Air Compressors.

T61.003B LP-B-20, Obj. D. Given a set of plant conditions or parameters indicating a partial or total loss of Instrument Air, Identify the correct procedure to be utilized and Outline the high level actions to stabilize the plant.

Question Source: Bank # ___R14982___

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(CFR: 43.5)

Comments:

K/A Match: Candidate must understand that the C Air compressor is cooled using the Chilled Water system which is no longer supplied by the chillers and they will trip (shutdown) on high temperature due to no cooling water.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 W/E04 LOCA Outside Containment / 3 Group # 1 K/A # EA2.1 Importance Rating 4.3 Ability to determine and interpret the following as they apply to the (LOCA Outside Containment): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Question #81 The crew is completing the actions of ECA-1.2, LOCA Outside Containment.

The following plant conditions exist:

• Pressurizer level is 35% and rising.

• Steam Generator narrow range levels are 30%.

• RCS Subcooling is 35ºF.

• RCS pressure is 1400 psig and rising.

The CRS will direct transition to which of the following procedure?

A. E-0, Reactor Trip or Safety Injection B. E-1, Loss of Reactor or Secondary Coolant C. ECA-1.1, Loss of Emergency Coolant Recirculation D. ES-1.2, Post LOCA Cooldown and Depressurization Answer: B Explanation:

A. Incorrect - E-0 directs a transition to ECA 1.2 in the RNO of E-0 step 20. There is no transition back to E-0 directed in ECA-1.2.

B. Correct - step 3b of ECA 1.2, direct the operator to transition to E-1 since step 3.a checks to see if RCS pressure is rising (indication that the LOCA outside Containment has been isolated)

C. Incorrect - the RNO is only performed if RCS Pressure is Lowering D. Incorrect - Transition to ES-1.2 is only transitioned to from E-1 or ES-1.1. All conditions are met to enter ES-1.2, however this transition is only made from ES-1.1 after completing actions of E-1.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Technical Reference(s):

1. ECA-1.2 LOCA outside Containment Rev 7,

2. BD-ECA- 1.2 Rev 0

3. E-1, Loss of Reactor or Secondary Coolant Rev 17 References to be provided to applicants during examination: None Learning Objective: T61.0003D Emergency Operations LP # D-14, Objective E - STATE and EXPLAIN the parameters which are evaluated, including their Criteria and Basis, to transition from ECA-1.2 to other procedures Question Source: Bank # __X L14389____

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2005_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

10 CFR: 55.43(5) - Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 033 Loss of Intermediate Range Nuclear Group # 2 Instrumentation K/A # AA2.09 Importance Rating 3.7 Ability to determine and interpret the following as they apply to the Loss of Intermediate Range Nuclear Instrumentation: Conditions which allow bypass of an intermediate-range level trip switch.

Question #82 Given the following plant conditions:

• A Reactor Startup is in progress

• The turbine is being prepared for loading with reactor power at approximately 8%

• Annunciator 77B, IR HI VOLT FAIL, alarms

• It is determined that Intermediate Range channel N36 has failed LOW What actions are required?

A. Trip bistables for channel N36 AND raise power above 10%

B. Trip bistables for channel N36 AND lower power below 1 x 10-8 amps C. Place the N36 level trip switch in the bypass position AND raise power above 10%

D. Place the N36 level trip switch in the bypass position AND lower power below 1 x 10-8 amps Answer: C Explanation:

T/S 3.3.1 Condition F is the applicable T/S for this malfunction. Per this action statement power must be raised above P-10 or lowered below P-6. OTO-SE-00001, Nuclear Instrumentation Malfunction, for this malfunction after power has been adjusted, the malfunctioning IR channel is bypassed by placing the level trip switch in Bypass.

A. Incorrect. In this condition if the bistable for N36 is tripped the reactor will trip. This is a plausible distractor because it is the action to take for a similar malfunction for a Power Range NI, and this is the correct power level to raise power to in this condition.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator B. Incorrect. In this condition if the bistable for N36 is tripped the reactor will trip. This is a plausible distractor because it is the action to take for a similar malfunction for a Power Range NI, however this is the incorrect power level to lower power to in this condition. This value is a

-10 common value used to take critical data. Lowering below the P-6 setpoint of 1x10 amps is a correct action for this failure.

C. Correct. This is the correct action and power level for this condition per T/S 3.3.1 condition F, and OTO-SE-00001, Nuclear Instrument Malfunction.

D. Incorrect. This is a correct action per OTO-SE-00001, however this is the incorrect power level to lower power to in this condition. This value is a common value used to take critical data.

-10 Lowering below the P-6 setpoint of 1x10 amps is a correct action for this failure.

Technical Reference(s): T/S 3.3.1, Condition F, and OTO-SE-00001, Nuclear Instrument Malfunction Rev 24.

References to be provided to applicants during examination: None.

Learning Objective: T61.003B, LP-B-42, Obj. D. Given a set of plant conditions or parameters indicting a Nuclear Instrument Malfunction, Analyze the correct procedure to be utilized and the required actions to stabilize the plant.

T61.0110 LP-28, Obj. H. State the LCOs associated with the following Nuclear Instrumentation Technical Specifications: 3.3.1, Table 3.3-1.

Question Source: Bank # ______

Modified Bank # ___L14017___

New _______

Question History: Last NRC Exam ______N/A______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 43.2)

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 000036 (BW/A08) Fuel Handling Group # 2 Accident / 8 K/A # 2.4.41 Importance Rating 4.6 Knowledge of the emergency action level thresholds and classifications.

Question #83 The plant is at 100% power.

Irradiated Fuel moves are in progress in the Spent Fuel Pool.

An Earthquake is felt at the site and results in the following:

• ANN 98E, Seismic Recorder On, is LIT

• HI HI RAD alarms are received on Fuel Building Process Radiation Monitors, GG-RE-27 and GG-RE-28, gas channels

• ANN 76D, SFP LEV HILO, is LIT

• ANN 62B, AREA RAD HI, is LIT Normal makeup to the SFP is initiated and level stabilizes at -20.0".

A report from the Fuel Building indicates that a fuel move was in progress when the earthquake occurred resulting in several pieces of fuel handling equipment and the assembly being moved, falling and damaging other fuel assemblies.

Area Radiation Monitors readings are as follows:

• SDRE0033, CTRL RM RAD MONITOR, is 1 mr/hr and constant

• SDRE0037, FUEL POOL BRIDGE CRANE RAD MONITOR, is 35 mr/hr and rising slowly

• SDRE0038, SPENT FUEL POOL AREA RAD MONITOR, is 32 mr/hr and rising slowly Which of the following describes the HIGHEST Emergency Plan Action Level that applies to this situation and why?

A. Unusual Event due to the Seismic Event B. Unusual Event due to Spent Fuel Pool Low Level C. Alert due to Area Dose Rates D. Alert due to Damage to Irradiated Fuel Assemblies

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Answer: D Explanation:

A. Incorrect - while there is an UE due to Ann 98E on and the earthquake felt at the site, it is NOT the highest EAL that exists B. Incorrect - while there is an UE due to low SFP level (less than -4 inches), it is NOT the highest EAL that exists C. Incorrect - Control Room Dose rates are 1 mr/hr which is less than the EAL threshold of 15mr/hr.

D. Correct - GGRE27 and 28 are in HI HI alarm resulting in an Alert (RA2.1) due to damage to irradiated fuel assemblies.

Technical Reference(s):

1. OTO-KE-00001, Fuel Handling Accident Rev 14

2. EIP-ZZ-00101 ADD 1, Emergency Action Level Classification Matrix, Rev 3

3. Annunciator 62B, Area Radiation High, Rev, 1

4. Annunciator 76D, Spent Fuel Pool Level High / Low, Rev, 1

5. Annunciator 98E, Seismic Recorder On, Rev 2 References to be provided to applicants during examination:

1. EIP-ZZ-00101 ADD 1, Emergency Action Level Classification Matrix, Rev 3 Learning Objective:

From T61.0110, Systems, LP #76, (SRO-RER-3), Objective C Identify initial classifications and potential escalations when plant conditions change.

From T61.0003D, Emergency Operations, LP RERP (SRO-AEO-1 and SRO-RER-2) - Given a set of plant conditions, classify an event using EIP-ZZ-00101 Question Source: Bank # ______

Modified Bank # ______

New __X_____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator 10 CFR 55.43.5 - Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

10 CFR 55.43.4 - Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 W/E15 Containment Flooding / 5 Group # 2 K/A # 2.4.18 Importance Rating 4.0 Knowledge of the specific bases for EOPs.

Question #84 Given the following plant conditions:

• A Reactor Trip and Safety Injection have occurred due to LOCA.

• ES-1.3, Transfer to Cold Leg Recirculation, has been completed.

The STA reports the following conditions:

• Containment normal sump level = 128 inches

• Maximum Containment Pressure reached during the accident was 20 psig The Crew has transitioned to FR-Z.2, Response to Containment Flooding,

• A CCW Leak in Containment has been identified and isolated.

• Step 2 is currently being performed to; Check IF Containment Sump Activity Level Can Be Determined.

What is the Bases of this Step?

A. To verify containment radiation levels.

B. To confirm the unexpected source of water.

C. To perform an offsite Dose Assessment per the RERP plan.

D. To determine if a transfer of containment sump water to plant storage tanks outside containment is possible.

Answer: D Explanation: The first sentence of Basis of this step states The step instructs the operator to determine the activity level in the containment sump water in order to provide information concerning the possible transfer of containment sump water to plant storage tanks outside the containment.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator A. Incorrect - The CHARMs provide the indication of containment radiation levels. The student may choose this is they think that a secondary verification of containment rad levels is needed.

This is not the basis of FR-Z.2. Containment High radiation is covered in FR-Z.3, Response to High Containment Radiation Levels.

B. Incorrect - This is the bases of step #1 of FR-Z.2.

C. Incorrect - RERP plan utilizes gaseous activity release values to perform Dose Assessments.

The was no indication that a release has occurred.

D. Correct per explanation above Technical Reference(s):

1. BD-FR-Z.2 Bases Document for Response to Containment Flooding, Rev 001

2. FR-Z.2 Response to Containment Flooding, Rev 007 References to be provided to applicants during examination: None Learning Objective: None - T61.003D, Emergency Operations, LP #30 Question Source: Bank # ______

Modified Bank # ______

New ____X___

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(10 CFR 55 .43.4)

Comments:

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NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 CE/A11; W/E08 RCS Overcooling - PTS Group # 2

/4 K/A # EA2.2 Importance Rating 4.1 Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock): Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

Question #85 Given the following plant conditions:

• The crew has performed E-0, Reactor Trip or Safety Injection.

• Upon transition to E-1, Loss of Reactor or Secondary Coolant, the following conditions exist:

o RCS temperature is 170°F and lowering slowly.

o RCS pressure is 50 psig and stable.

o RHR flow indicates 6000 GPM.

Which of the following describes the status of the RCS INTEGRITY Critical Safety Function Status Tree, and the NEXT action required upon exit from E-0?

REFERENCE PROVIDED A. An ORANGE condition exists on the RCS INTEGRITY Critical Safety Function Status Tree. Verify RHR flow rate in FR-P.1 and then transition back to E-1 due to the Large Break LOCA in progress.

B. An ORANGE condition exists on the RCS INTEGRITY Critical Safety Function Status Tree. Reduce RHR flow and initiate an RCS temperature soak in accordance with FR-P.1, Response to Imminent Pressurized Thermal Shock Conditions.

C. A RED condition exists on the RCS INTEGRITY Critical Safety Function Status Tree. Verify RHR flow rate in FR-P.1 and then transition back to E-1 due to the Large Break LOCA in progress.

D. A RED condition exists on the RCS INTEGRITY Critical Safety Function Status Tree. Reduce RHR flow and initiate an RCS temperature soak in accordance with FR-P.1, Response to Imminent Pressurized Thermal Shock Conditions.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Answer: C Explanation: When RCS Pressure and Temperature are plotted on Figure 4a of CSF-1 (page 7 of 11), the values are to the Left of the Limit A Curve When checking the severity of the Integrity CSF (Figure 4 of CSF-1, page 6 of 11) Answer NO to the question Temperature Reduction in ALL RCS Cold Leg Less than 100F in 60 minutes and Answer NO to question ALL RCS Pressure VS Cold Leg Temperature Points to the Right of Limit A. This results in a RED Path.

For step #1 of FR-P.1, Check RCS Pressure - Greater Than 325 psig, the answer is NO. The RNO states IF either RHR pump flow is greater than 850 GPM, THEN Return To procedure and step in effect. Therefore RHR flow is left as is and a Transition back to E-1 is performed.

A. Incorrect a RED path exists B. Incorrect a RED path exists C. Correct - see above D. Incorrect - RHR is not reduced IAW FR-P.1.

Technical Reference(s):

1. CSF-1, Critical Safety Function Status Trees (CSFST), Rev 10

2. FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, Rev 10 References to be provided to applicants during examination: CSF-1, Figure 4a.

Learning Objective:

T61.003D, Emergency Operations, LP #28 Objective G STATE and EXPLAIN the parameters which are evaluated, including their Criteria and Basis, to transition from the following procedures to another procedure.

1. FR-P.1, Response To Imminent Pressurized Thermal Shock Condition.

Question Source: Bank # __X L16706____

Modified Bank # ______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

(10 CFR 55.43.5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

This question cannot be answered solely using fundamental knowledge of plant parameters that require direct entry to major EOPs; e.g., major Westinghouse EOPs are E0, E1, E2, E3, ECA-0.0, and Red/Orange Functional Restoration Procedures and major General Electric EOPs are

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Reactor Vessel Control, Primary Containment Control, Secondary Containment Control, and Radioactive Release Control.

To correctly answer this question the student must be able to perform the following:

1. Correctly use the reference to determine that a RED Path exists Then

2. Have knowledge of specific requirements of the first step of FR-P.1 that requires the SRO to transition back to the procedure and step in effect. This is NOT a direct entry into an EOP.

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 005 Residual Heat Removal Group # 1 K/A # A2.01 Importance Rating 2.9 Ability to (a) predict the impacts of the following malfunctions or operations on the RHRS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure modes for pressure, flow, pump motor amps, motor temperature, and tank level instrumentation Question #86 Given the following plant conditions:

• The Plant is in Mode 6 with > 23 feet above the top of the vessel flange.

• Residual Heat Removal Train B is in service in a cooldown lineup.

• A RHR Pump is Out of Service for breaker replacement.

• EJ FK-619, RHR HX B BYPASS CTRL, is in AUTO controlling flow at 3500 gpm as indicated on EJ FI-619, RHR TO ACC INJ LOOPS 3 & 4 FLOW.

• EJ HCV-607, RHR HX B FLOW CTRL VLV is 15% open.

Subsequently, EJ FT-611, RHR Pump B Recirc flow transmitter fails LOW.

(1) Which of the following describes the system response?

And (2) In accordance with Technical Specifications, 1000 gpm RHR flow to the RCS should be verified in order to _______.

A. (1) EJ FCV-619, RHR HX B Bypass Flow Control Valve, closes.

(2) prevent boron stratification.

B. (1) EJ FCV-611, B RHR Recirc Valve, opens.

(2) prevent boron stratification.

C. (1) EJ FCV-619, RHR HX B Bypass Flow Control Valve, closes.

(2) protect the RHR Pump impeller from overheating.

D. (1) EJ FCV-611, B RHR Recirc Valve, opens.

(2) protect the RHR Pump impeller from overheating.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Answer: B Explanation: Per the Technical Specification Bases of 3.9.5 . The flow rate is determined by the flow rate necessary to provide sufficient decay heat removal capability and to prevent thermal and boron stratification in the core.

Per the M22 print FCV619 is control by the output of FT 619 and will open to compensate for the false low system flow coming from this controller. The RHR pump Recirc valve is controlled by a different FE, FE611, and therefore would not see a false low flow and would remain closed in this situation. There is no TRUE low flow condition in fact total system flow is high, RHR pump overheating is not a concern but runout would be a concern.

A. Incorrect. EJ FCV-619 will respond to the reduced flow through the system by OPENING to raise flow out of the RHR system.

B. Correct. EJ FCV-611 will open when flow is reduced to approximately 783 to 816 gpm, temperature dependent, and recirculate flow to the suction of the RHR pump at approximately 600 gpm.

C. Incorrect EJ FCV-619 will respond to the reduced flow through the system by OPENING to raise flow out of the RHR system. The function of the recirc valve EJ FK-611 is to maintain flow through the system to prevent damage to the pump from overheating or vibration when running at shutoff head or reduced flow conditions, however the basis of verifying 1000 gpm flow through the RHR train to the RCS is to prevent thermal and boron stratification in the core during shutdown conditions.

D. Incorrect The function of the recirc valve EJ FK-611 is to maintain flow through the system to prevent damage to the pump from overheating or vibration when running at shutoff head or reduced flow conditions, however the basis of verifying 1000 gpm flow through the RHR train to the RCS is to prevent thermal and boron stratification in the core during shutdown conditions.

Technical Reference(s):

1. Technical Specification 3.9.5 and its bases

2. M-22EJ01(Q), Revision 59, Piping and Instrument Diagram RHR System

3. OTN-EJ-00001, RHR system, Rev 28 References to be provided to applicants during examination: None Learning Objective:

1. T61.0110 Systems, LP#7 Residual Heat Removal, Objective B DESCRIBE the purpose and operation of the following RHR System components, to include interlocks, controller operation and power supplies.

1. RHR pumps

2. RHR Heat Exchangers (HXs)

3. Reactor Coolant System (RCS) Hot Leg Suction Valves to RHR

4. Containment (CTMT) recirculation (Recirc) Sump Suction Valves to RHR

5. Refueling Water Storage Tank (RWST) Suction Valves to RHR

6. RHR Heat Exchanger Flow Control Valves

7. RHR Heat Exchanger Bypass Valves

8. RHR Bypass Miniflow Valves

9. RHR Discharge Valves to the Safety Injection (SI) and Centrifugal

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Charging Pumps

10. Suction Relief Valves

11. RHR Cold Leg Injection Valves

12. RHR SI System Hot Leg Recirc Isolation Valves

13. RHR Hot Leg Injection Isolation Valve

14. RHR Downstream Relief Valves Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

10 CFR:55.43.2 - Facility operating limitations in the Technical Specifications and their bases.

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 006 Emergency Core Cooling Group # 1 K/A # A2.02 Importance Rating 4.3 Ability to (a) predict the impacts of the following malfunctions or operations on the ECCS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of flow path.

Question #87 Given the following plant conditions:

• The plant is at 100% power

• An electrical short causes EJ HV8716A, RHR TRN A SI SYS HOT LEG RECIRC ISO valve to CLOSE (1) What is the status of the B Train RHR system, And (2) What action is required for this failure?

A. (1) Operable (2) Verify 100% equivalent ECCS flow on B train RHR B. (1) Operable (2) Restore system lineup IAW OTN-EJ-00001, RHR System C. (1) Inoperable (2) Place the plant in MODE 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> D. (1) Inoperable (2) Open EJ HV8716B, RHR TRN B SI SYS HOT LEG RECIRC ISO valve Answer: C Explanation:

A. Incorrect. Plausible because the valve is associated with A train, however to maintain both trains operable, each train must be able to inject into all four loop cold legs. When this valve closes each train is only capable of injecting to two cold legs making both trains Inoperable. If the candidate assumes that only A train is effected, T/S 3.5.2 for ECCS operating verifies an

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator equivalent of 100% flow capability on the opposite train of that component is available, and then directs the crew to restore the system to operable status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

B. Incorrect. Plausible because the valve is associated with A train, however to maintain both trains operable, each train must be able to inject into all four loop cold legs. When this valve closes each train is only capable of injecting to two cold legs making both trains Inoperable. If other RHR system valves had failed to a position other than specified for standby lineup, then they would be restored to standby lineup in accordance with OTN-EJ-00001, Add 5 or 6 for trains A and B.

C. Correct. To maintain both trains operable, each train must be able to inject into all four loop cold legs. When this valve closes each train is only capable of injecting to two cold legs making both trains Inoperable With both trains inoperable entry into T/S 3.0.3 is made and the plant must be placed in Mode 3 within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

D. Incorrect. Plausible because B train is also inoperable, and if it is believed that the two cross connect valves are in parallel and not in series, then opening EJ HV8716B would restore the cross connect capability and therefore operability.

Technical Reference(s):

1. OTN-EJ-00001, RHR System, Rev 28,

2. T/S 3.5.2, ECCS Operating,

3. T/S 3.0.3, ODP-ZZ-00002 Attachment 3 References to be provided to applicants during examination: None Learning Objective: T61.0110 LP-07, Obj. F. State the LCOs for the ECCS trains Technical Specifications and Discuss the RHR Systems function as it pertains to these Technical Specifications.

Question Source: Bank # ______

Modified Bank # ______

New ____X___

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR: 43.5)

Comments:

K/A Match: Candidate must apply knowledge that the valve closure reduces the flow of each train of RHR to less than the required four loops for its ECCS function, and then apply that knowledge to the T/S actions required for the inoperable portion of the system.

SRO- Candidate must apply T/S 3.0.3 for both trains of RHR being inoperable and not able to perform the function of ECCS T/S 3.5.2, ECCS Operating.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 013 Engineered Safety Features Group # 1 Actuation K/A # 2.1.7 Importance Rating 4.7 Ability to evaluate plant performance and make operational judgments based on operating characteristics, reactor behavior, and instrument interpretation.

Question #88 Given the following plant conditions:

• The plant is at 100% power

• Channel 2 of the ESFAS Cabinet is being de-energized in accordance with Attachment 4, De-Energizing Channel II ESFAS Cabinets, of OTS-SA-00001, Operation of Engineered Safety Feature Actuation System

• Annunciator 128B, TD AFP START, alarms (1) Is this an expected alarm?

And (2) Which of the following functions is made INOPERABLE by de-energizing Channel 2 ESFAS?

A. (1) Yes (2) Turbine Trip on SG high water level.

B. (1) Yes (2) Auxiliary Feedwater Pump Suction Transfer on low suction pressure.

C. (1) No (2) Turbine Trip on SG high water level.

D. (1) No (2) Auxiliary Feedwater Pump Suction Transfer on low suction pressure.

Answer: B Explanation: Per Attachment 4, De-Energizing Channel II ESFAS Cabinets, of OTS-SA-00001, Operation of Engineered Safety Feature Actuation System Annunciator 128B, TD

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator AFP START, will alarm.

Per OTS-SA-00001, Operation of Engineered Safety Feature Actuation System, De-energizing only make the AFAS Manual Initiation and Auxiliary Feedwater Pump Suction Transfer on low suction pressure functions INOPERABLE.

A. Incorrect - This alarm will actuate, however the Turbine Trip on SG high water level function is INOPERABLE if Channel 1 or 4 is de-energized B. Correct - This alarm will actuate, Per OTS-SA-00001, Operation of Engineered Safety Feature Actuation System, De-energizing only make the AFAS Manual Initiation and Auxiliary Feedwater Pump Suction Transfer on low suction pressure functions INOPERABLE.

C. Incorrect - This alarm will actuate however the Turbine Trip on SG high water level function is INOPERABLE if Channel 1 or 4 is de-energized D. Incorrect - This alarm will actuate Per OTS-SA-00001, Operation of Engineered Safety Feature Actuation System, De-energizing only make the AFAS Manual Initiation and Auxiliary Feedwater Pump Suction Transfer on low suction pressure functions INOPERABLE.

Technical Reference(s):

OTO-NB-00001, Loss of Power to NB01, Rev 18 References to be provided to applicants during examination: None Learning Objective:

T61.0110 Systems, LP #52 EFSAS, Objective F. DISCUSS the purpose and scope of the following:

1. OTS-SA-00001, "De-energizing and Energizing Engineered Safety Feature Actuation System.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

10CFR 55.43.2 Facility operating limitation and Technical Specifications and their bases.

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 062 AC Electrical Distribution Group # 1 K/A # A2.10 Importance Rating 3.3 Ability to (a) predict the impacts of the following malfunctions or operations on the ac distribution system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations:

Effects of switching power supplies on instruments and controls Question #89 The plant is at 100% power:

• Annunciator 28B, NN14 INV TRBL/XFR, has just alarmed.

• Local indication shows that the Bypass Source Supplying Load, P202, light is LIT.

1) What is the status of NN14, 120V AC Vital Inverter?

And 2.) What Technical Specification action(s) will to be taken to mitigate the situation?

____(1)____ ____(2)____

A. INOPERABLE 3.8.7 Inverters Operating ONLY B. OPERABLE 3.8.7 Inverters Operating ONLY C. INOPERABLE 3.8.7 Inverters Operating AND 3.8.9 Distribution Systems Operating D. OPERABLE 3.8.7 Inverters Operating AND 3.8.9 Distribution Systems Operating Answer: A Explanation: Per T/S 3.8.7, in order to be considered operable the inverters be powered

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator from a 125 VDC station battery, and supply power to the associated Vital bus.

A. Correct -NN14 is INOPERABLE until it is supplying the 120V AC Vital Bus NN04 from the battery. NN04 will remain OPERABLE if it is energized from the bypass source per T/S 3.8.9.

T/S 3.8.7 for the INOPERABLE inverter is the only T/S entered.

B. Incorrect - The inverter NN14 is INOPERABLE, however NN04 will remain OPERABLE if it is energized from the bypass source. T/S 3.8.7 for the INOPERABLE inverter is the only T/S entered.

C. Incorrect - NN14 is INOPERABLE until it is supplying the 120V AC Vital Bus NN04 from the battery. NN04 will remain OPERABLE if it is energized from the bypass source per T/S 3.8.9.

T/S 3.8.7 for the INOPERABLE inverter is the only T/S entered. T/S 3.8.9 is referenced from T/S 3.8.7 to be entered ONLY if NN04 is deenergized. In this case NN04 is energized from the bypass source.

D. Incorrect - NN14 is INOPERABLE until it is supplying the 120V AC Vital Bus NN04. NN04 will remain OPERABLE if it is energized from the bypass source. T/S 3.8.7 for the INOPERABLE inverter is the only T/S entered. T/S 3.8.9 is referenced from T/S 3.8.7 to be entered if NN04 is deenergized. In this case NN04 is energized from the bypass source.

Technical Reference(s): TS 3.8.7 Bases References to be provided to applicants during examination: None Learning Objective: Safeguards Power G. EXPLAIN the Technical Specifications and bases for the Safeguards Power System.

Question Source: Bank # ______

Modified Bank # ______

New __X_____

Question History: Last NRC Exam __N/A__________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: SRO 2 (CFR: 43..2)

Comments: Based of a question on the 2010 CPNPP SRO exam

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 003 Reactor Coolant Pump Group # 1 K/A # 2.4.6 Importance Rating 4.7 Knowledge of EOP mitigation strategies.

Question #90 Given the following plant conditions:

• A LOCA has occurred at the weld connection of the Pressurizer Spray line connection at the top of the Pressurizer.

• A Reactor trip and Safety Injection have occured.

• The Crew is performing step 9 of E-1, Loss of Reactor or Secondary Coolant.

• The BOP reports All S/G pressures are 950 psig and slowly Lowering.

• RCS Pressure is 1350 psig and stable.

• Containment Radiation Monitors indicate the following:

o GT-RE-59, is reading 8x103 R/hr.

o GT-RE-60, is reading 1x104 R/hr.

(1) What is the status of the Steam Generator pressure?

And (2) What is the Emergency Action Level associated with this event?

A. (1) Steam Generators are Faulted (2) Alert B. (1) Steam Generator Pressures are Stable (2) Alert C. (1) Steam Generators are Faulted (2) Site Area Emergency D. (1) Steam Generator Pressures are Stable (2) Site Area Emergency Answer: D

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Explanation: For the EAL Determination, 2 Fission product barriers are lost: RCS system Barrier and Fuel Clad Barrier. When the candidate applies the matrix (Table F-1), they will determine that a FS1.1 exists: Site Area Emergency.

Per the E-1, Loss of Reactor or Secondary Coolant background document for Step 9, Check SG and RCS Pressures: With a LOCA and no faulted SG the SG pressure could be lowering slightly.

This is considered a "stable" SG pressure. The concern addressed by this step is the presence of a secondary side break in which the faulted SG is still depressurizing in an uncontrolled manner.

If this is the case, the SI termination criteria may not be met at the time the check is encountered, and the operator should return to Step 1 in E-1 and not proceed to ES-1.2, Post LOCA Cooldown And Depressurization, until all SG pressures have been stabilized or are rising and RCS pressure has stabilized or is lowering.

A. Incorrect - both portions are incorrect. Plausible if the student assumes that lowering S/G pressure is indicative of a Faulted S/G during this procedure.

B. Incorrect - The EAL is incorrect. Plausible if the candidate does not apply the GT RE-60 reading correctly to the criteria for a barrier loss.

C. Incorrect - Steam Generators are NOT considered Faulted.

D. Correct - See Explanation Above.

Technical Reference(s):

1. EIP-ZZ-00101 Addendum 1, EAL Classification Matrix, Rev 3

2. E-1 Background Document, REV 9 References to be provided to applicants during examination:

1. EIP-ZZ-00101 Addendum 1, EAL Classification Matrix, Rev 3 Learning Objective:

1. T61.003D Emergency Operations - LP #D-8, Obj. A. Explain the purpose and Major Action Categories of E-1, loss of Reactor or Secondary Coolant.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

10 CFR:55.43.1 - conditions and limitations in the facility license.

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 014 Rod Position Indication Group # 2 K/A # 2.2.38 Importance Rating 4.5 Knowledge of conditions and limitations in the facility license.

Question #91 Given the following plant conditions:

• Mode 1 at 100% Rated Thermal power.

• RPI NON-URG annunciator, Ann 80B, is LIT.

• DRPI Data 'A' Failure LED is LIT.

• DRPI Data B Failure LES is NOT LIT.

• Rod Bottom Light for H-8 is NOT LIT.

• NO other Rod Control System alarms are LIT.

Which of the following describes the operability of DRPI and the accuracy of determining the position of rod H-8?

A. Operable; accuracy is +10, -4 steps B. Inoperable; accuracy is +10, -4 steps C. Operable; accuracy is +/- 4 steps D. Inoperable; no measureable accuracy Answer: A Explanation:

Per the Annunciator 80B, From the display panel, the range of any lighted LED at half-accuracy under "Error in A" condition is +10, -4 steps. Also per the technical specification 3.1.7 bases, To increase the reliability of the system, the inductive coils are connected alternately to data system A or B. Thus, if one data system fails, the DRPI will go on half accuracy. The DRPI system is capable of monitoring rod position within at least +/- 12 steps with either full accuracy or half accuracy. (Background section - last paragraph, page B 3.1.7-2).

Background - two separate and independent systems: the Bank Demand Position Indication System (commonly called group step counters) and the Digital Rod Position Indication (DRPI)

System. Based on the indication present, Annunciator 80C and Annunciator 80A NOT LIT, and Rod Bottom Light NOT LIT, the student can determine that the Demand Position Indication System is working properly and the malfunction is only associated with DRPI system.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Per the LCO section of the Tech Spec Bases: Indication System be OPERABLE for each control and shutdown rod. For the rod position indicators to be OPERABLE requires meeting the SR of the LCO and the following:

a. The DRPI System, on either full accuracy or half accuracy, indicates within 12 steps of the group step counter demand position as required by LCO 3.1.4, "Rod Group Alignment Limits";

AND

b. The Bank Demand Indication System has been calibrated either in the fully inserted position or to the DRPI System.

Therefore the system is Operable.

A. Correct - see above B. Incorrect - it is operable C. Incorrect the accuracy per Annunciator 80B is +10, -4 steps D. Incorrect - it is operable Technical Reference(s):

1. Technical Specifications Bases for 3.1.7, Rod Position Indication

2. OTA-RK-00022 ADD 80B,RPI Non Urgent Alarm, Rev 5

3. OTA-RK-00022 ADD 80C,RPI Rod Deviation, Rev 5

4. OTA-RK-00022 ADD 80A,RPI Urgent Alarm, Rev 2 References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems, LP 26 Rod Control Objective S: EXPLAIN how DRPI full and half accuracy are developed.

Objective U: STATE the Technical Specification limiting conditions for operations (LCOs) applicable to the rod control system.

Question Source: Bank # __X R8566____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

CFR: 43.1 - Conditions and limitations in the facility license.

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 033 Spent Fuel Pool Cooling Group # 2 K/A # A2.01 Importance Rating 3.5 Ability to (a) predict the impacts of the following malfunctions or operations on the Spent Fuel Pool Cooling System; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Inadequate SDM Question #92 Given the following plant conditions:

• The plant is operating at 100%

• B Train Spent Fuel Pool Cooling System is in service

• B CCW Train is in service

• A leak develops in the B Spent Fuel Pool Cooling Heat Exchanger (1) What impact will this have on the Spent Fuel Pool, And (2) Which procedure will be entered to mitigate the consequences of this failure?

A. (1) Spent Fuel Pool LEVEL will LOWER (2) OTN-EG-00001, Component Cooling Water System B. (1) Spent Fuel Pool Boron Concentration will LOWER (2) OTN-EG-00001, Component Cooling Water System C. (1) Spent Fuel Pool LEVEL will LOWER (2) OTN-EC-00001, Fuel Pool Cooling and Cleanup System D. (1) Spent Fuel Pool Boron Concentration will LOWER (2) OTN-EC-00001, Fuel Pool Cooling and Cleanup System Answer: D Explanation:

A. Incorrect. Plausible because the student could incorrectly assume that a leak in the SFP HX will result in a loss of SFP level. See Explanation D. SFP level will rise not lower. OTN-EG-00001 does not contain information on how to mitigate the consequences of this failure.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator B. Incorrect. Plausible because SFP boron concentration will lower, and incorrect because OTN-EG-00001 does not contain information on how to mitigate the consequences of this failure.

C. Incorrect. Plausible because the student could incorrectly assume that a leak in the SFP HX will result in a loss of SFP level. See Explanation D. OTN-EC-00001 has direction to refill the spent fuel pool under normal losses such as evaporation.

D. Correct. CCW system pressure in the SFP HX is higher than the operating pressure of the SFP Cleanup system, therefore CCW will leak into the SFP system causing a dilution of the SFP boron concentration reducing SDM. In response to the Rising level in the SFP and reduced boron concentration, OTA-RK-00022, ADD 76D, SFP Level HI/LO will direct entry to OTN-EC-00001 to Divert water from the SFP to either the RWST or RHUT, and direct boron sampling and adding boron to the SFP if required.

Technical Reference(s): OTN-EC-00001, Fuel Pool Cooling and Cleanup System, REV 39, OTA-RK-00022, ADD 76D, Spent Fuel Pool Level HIGH/LOW, REV 1, OTO-EC-00001, Loss of SFP/Refuel Pool Level, REV 12.

References to be provided to applicants during examination: None.

Learning Objective: T61.0110 LP-24, Obj. I Explain the precautions and limitations and bases for the following conditions/processes associated with OTN-EC-00001, Fuel Pool Cooling and Cleanup System 3. Minimum boron concentration, Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

10 CFR Part 55 Content:

(CFR: 43.5)

Comments:

K/A Match: Candidate must assess the effects of a CCW leak in the SFP HX and determine that a dilution event is in progress reducing the SDM of the SFP, and then select the appropriate recovery procedure to restore the SFP boron concentration.

SRO ONLY: Candidate must make procedure selection to address the event. The question CANNOT be answered solely by knowing entry conditions for AOPs

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 034 Fuel Handling Equipment System (FHES) Group # 2 K/A # K4.02 Importance Rating 3.3 Knowledge of design feature(s) and/or interlock(s) which provide for the following: Fuel movement.

Question #93 Given the following plant conditions:

• The plant is in Mode 6 during refueling.

• A fuel assembly is raised to just clear the top of the Upender.

• The operator tries to move the trolley to take the fuel assembly to its new location in the core but the trolley does NOT move.

What action must be performed to move the trolley?

A. Move the bridge to the new location B. Raise the hoist to its hoist up position C. Rotate the mast until the mast rotation light is out D. Bypass the interlock by depressing the Load Bypass pushbutton Answer: B Explanation:

A. Incorrect -. OTS-KE-00013 specifies the interlocks associated with the Bridge stating that Bridging right/left with the hoist less than full up and loaded is limited to specific areas, including a 2 inch square area above the Upender. If the bridge is near the edge of the 2 inch square boundary, the bridge will not move until the hoist in the hoist up position, However bridge relocation itself will not allow the trolley to be moved.

B. Correct - OTS-KE-00013 specifies the interlocks associated with the Trolley stating that Trolley forward/reverse with the hoist less than full up and loaded is limited to specific areas, including a 2 inch square area above the Upender. If the trolley is near the edge of the 2 inch square boundary, the trolley will not move until the hoist in the hoist up position.

C. Incorrect -.Plausible because the Mast being within 6 inches of the Temporary Zone will cause the Bridge and Trolley speed to slow to 3 fpm, and will stop bridge or trolley movement when at the boundary of the zone, however rotation of the Mast is not interlocked with Trolley movement.

D. Incorrect - Plausible because the interlock preventing movement of the Trolley outside the

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator specified zones with the hoist not in the hoist up position can be bypassed using the Interlock Bypass Key Switch, which is a different interlock bypass than the Load Bypass pushbutton which allows the HOIST to be lowered when a slack cable condition occurs.

Technical Reference(s): OTS-KE-00013, Refueling Machine, REV 31 References to be provided to applicants during examination: None Learning Objective: T61.003E, LP-E-5, Obj. E. Describe the function, capacity, and operation of the following equipment: 5. Refueling Machine.

Question Source: Bank # __F13344____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 43.7)

Comments: SRO #7

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Conduct of Operations Group # Generic K/A # 2.1.35 Importance Rating 3.9 Knowledge of the fuel-handling responsibilities of SROs.

Question #94 Who can authorize the bypassing of interlocks on the refueling machine console?

A. The Refueling SRO with concurrence of the Shift Manager.

B. The Refueling SRO with concurrence of the Reactor Engineering representative.

C. The Refueling SRO with concurrence from the Westinghouse Nuclear Fuels Division representative.

D. The Shift Manager with concurrence from the Westinghouse Nuclear Fuels Division representative.

Answer: A Explanation: Step 3.4 of OTS-KE-00013, states that When interlocks are bypassed, the system is being operated in an abnormal condition solely under the Refueling Senior Reactor Operator (SRO) judgment. The Refueling Senior Reactor Operator (SRO) judgment must be double-checked by a second Senior Reactor Operator (SRO) under these conditions. Keys for the interlock bypass switches shall be kept under administrative control.

A. Correct B. Incorrect - the Reactor Engineering representative is not an SRO licensed individual C. Incorrect - the Westinghouse Nuclear Fuel Division representative is not an SRO licensed individual. However at Callaway movement of nuclear fuel is contracted out to an off site organization, but is supervised by a Callaway Refueling SRO. It is plausible that the candidate could determine that offsite organizations involved in the fuel movement would have authority on fuel handling equipment.

D. Incorrect - the Westinghouse Nuclear Fuel Division representative is not an SRO licensed individual and while the Shift Managerr is a licensed individual and may provide concurrence, this individual is not directly in charge of the fuel moves and should not request to bypass the interlocks. This is the responsibility of the Refueling Supervisor. However at Callaway movement of nuclear fuel is contracted out to an off site organization, but is supervised by a Callaway Refueling SRO. It is plausible that the candidate could determine that offsite organizations involved in the fuel movement would have authority on fuel handling equipment.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Technical Reference(s):

1. OTS-KE-00013, Refueling Machine, Rev 31 References to be provided to applicants during examination: None Learning Objective: T61.003E License Refueling Training, LP#5 Fuel Handling systems, Objective H, Describe the interlocks and protective features of the following:

1. New fuel elevator

2. Spent fuel bridge crane

3. Transfer system

4. Refueling machine gripper Question Source: Bank # __X__F13338__

Modified Bank # ______

New _______

Question History: Last NRC Exam __N/A__________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

10 CFR 55.43.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Conduct of Operations Group # Generic K/A # 2.1.13 Importance Rating 3.2 Knowledge of facility requirements for controlling vital/controlled access Question #95 Which of the following personnel is required to be granted permission to enter the At the Controls Area of the Control Room?

A. NRC Resident Inspector B. On-Shift I&C Technician C. On-Coming Shift Manager D. Security Department Personnel Answer: D Explanation: Per ODP-ZZ-00001 Step 4.4.4.c The On-Shift/On-Coming Operations Crew and Resident NRC Inspectors may enter the "AT THE CONTROLS AREA" without obtaining permission from the SM/CRS. Additional permission is required to enter the "IMMEDIATELY ADJACENT TO THE CONTROL PANELS AREA". Radiation Protection Technicians, Radiation Protection Tech Support, RTOs, Chemistry Technicians and I&C Technicians are considered to be part of the On-Shift Operations Crew.

A. Incorrect - The NRC Resident Inspector does not require permission to enter the At the Controls area, but must obtain permission to enter the Immediately Adjacent to the Control Panels Area.

B. Incorrect - The On-Shift I&C technician is considered part of the On-Shift Operations Crew and does not require permission to enter the AT the Controls Area, however a Non-Shift technician would require permission.

C. Incorrect - The Oncoming SM does not require permission to enter for Turnover, however a SM who is not part of the On-Shift Crew or On-Coming Crew must still obtain permission to enter.

D. Correct - The Security department personnel do not require permission to enter the Control Room at any time, however to enter the At The Controls Area, the CRS/SM must grant permission to enter for all Security Personnel.

Technical Reference(s): ODP-ZZ-00001, Operation Department Code of conduct, Rev 91 References to be provided to applicants during examination: None

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Learning Objective: T61.0110 Systems, LP #66 Operations Department - Code of Conduct Objective A.2.a: EXPLAIN the following as applied in ODP-ZZ-00001, Operations Dept. - Code of Conduct: the responsibilities of the Shift Manager Question Source: Bank # ___L5963___

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 43.5)

Comments: SRO per #5, IT is the responsibility of the CRS/SM to grant permission to enter the At the Controls Area of the Control Room.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Equipment Control Group # Generic K/A # 2.2.20 Importance Rating 3.8 Knowledge of the process for managing troubleshooting activities.

Question #96 Due to an undetermined equipment deficiency, MDP-ZZ-TR001, Planning and Execution of Troubleshooting Activities, has been implemented.

It has been determined that the troubleshooting activity presents the risk of a Safety System actuation.

(1) Who is required to approve the Troubleshooting Plan?

And (2) The Troubleshooting Plan should be a _______ Plan?

A. (1) Director, Maintenance (2) Level 1 B. (1) Shift Manager (2) Level 1 C. (1) Director, Maintenance (2) Level 4 D. (1) Shift Manager (2) Level 4 Answer: B Explanation: Per the procedure MDP-ZZ-TR001, revision 16, Section 3.0 and Attachment 2 sheet 2 of 2, the Shift Manager and Duty Manager is the required approval authority.

Per Attachment 2 sheet 1 of 2, Troubleshooting Risk Level Assessment Guide, this is a level 1 activity as the question states that there is a risk of a safety system actuation. Therefore, it is a Level 1 because of the actuating a safety system, causing a significant reactivity change, and not a Level 4. Level 4 plans are Plant Equipment is REMOVED FROM SERVICE and Troubleshooting activities cannot affect the operation or safety of the plant.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator A. Incorrect - Wrong person for approval B. Correct as explained above C. Incorrect - Wrong level and approval person D. Incorrect - Wrong level Technical Reference(s):

1. MDP-ZZ-TR001, Planning and Execution of Troubleshooting Activities, Rev 16 References to be provided to applicants during examination: None Learning Objective: None Question Source: Bank # ______

Modified Bank # ______

New ____X___

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

CFR: 43.5 Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Equipment Control Group # Generic K/A # 2.2.5 Importance Rating 3.2 Knowledge of the process for making design or operating changes to the facility.

Question #97 Who has the authority to APPROVE or REJECT a proposed design change to safety related systems, structures or components (SSC's)?

A. Vice President, Nuclear B. Senior Plant Director C. Director, Nuclear Operations D. Shift Manager Answer: C Explanation: Step 4.1.7 of APA-ZZ-00600. Director, Nuclear Operations AND Director, Engineering Design or Director, Engineering Projects: APPROVE or REJECT the proposed design change to safety related SSCs.

A. Incorrect B. Incorrect C. Correct See explanation above D. Incorrect.

Technical Reference(s):

1. APA-ZZ-00600, Design Change Control, Revision 52 References to be provided to applicants during examination: None Learning Objective:

1. T61.003A, Normal Operations - MOD A, LP #18 Objective A PERFORM the following as it pertains to APA-ZZ-00600, Design Change Control

3. DISCUSS Concurrent Changes to include:

c. Operations approval requirements Question Source: Bank # __X L15426____

Modified Bank # ______

New _______

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

10 CFR Part 55 Content:

10 CFR: 55.43.3 Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Radiation Control Group # Generic K/A # 2.3.6 Importance Rating 3.8 Ability to approve release permits.

Question #98 A radwaste release from the Discharge Monitor Tank 'A' is in progress.

(1) Which of the following conditions would AUTOMATICALLY prevent the release?

And (2) If the release is delayed for more than 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, the original release permit may be considered valid if approval from whom is received?

A. (1) Cooling tower blowdown flow rate lowers to 7000 gpm (2) Director, Nuclear Operations B. (1) Cooling tower blowdown flow rate lowers to 7000 gpm (2) Manager, Radiation Protection C. (1) HB-RE-18, RW BLD Disch Line HB-RE-0018 Gamma Det, fails resulting in a Hi Hi alarm (2) Director, Nuclear Operations D. (1) HB-RE-18, RW BLD Disch Line HB-RE-0018 Gamma Det, fails resulting in a Hi Hi alarm (2) Manager, Radiation Protection Answer: D Explanation: The Approval must be granted by the Radiation Protection Supervisor or RPM (Manager, Radiation Protection) per step 4.2 of HTP-ZZ-02006 not the Director , Nuclear Operations. The low flow setpoint is adjustable but somewhere between 3000-5000 gpm. 7000 gpm is higher than these and therefore not correct as it would not prevent the release.

A. Incorrect as both (1) and (2) are incorrect B. Incorrect as the flow minimum is wrong

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator C. Incorrect as the approval authority is wrong D. Correct Technical Reference(s):

1. HSP-ZZ-00014, Rad Monitor Inoperability, Revision 25 - section 6.1

2. HTP-ZZ-00006, Liquid Radwaste Release Permit (Batch) Rev 84 References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems LP #16, Radwaste Objective E DESCRIBE the purpose and operation of the following Liquid Radwaste components / subsystems: #7 Liquid Release and Isolation Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _X____

Comprehension or Analysis _____

10 CFR Part 55 Content:

10 CFR 55.43.4 - Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Emergency Procedures / Plan Group # Generic K/A # 2.4.27 Importance Rating 3.9 Knowledge of fire in the plant procedures.

Question #99 Given the following plant conditions:

• The plant is operating at 100% power.

• A report is received by the Control Room at 1315 that a fire exists in the Lower Cable Spreading Room.

In accordance with EIP-ZZ-00101, Addendum 1, EAL Classification Matrix, what conditions will require a declaration of an UNUSUAL EVENT?

A. The fire is NOT extinguished by 1330.

B. The fire has caused EGHV0059, CCW to CTMT, to close.

C. The fire is now causing damage to the NK13 Battery Room.

D. Immediately following verification of an actual fire within the Protected Area.

Answer: A Explanation: The 15 minute time requirement specified in the Fire and Explosion EAL starts at the alarm or other indication of a fire and the clock stops when the alarm is verified to be spurious or the fire is verified out.

A. Correct B. Incorrect - The fire causing spurious operation of safety related equipment results in an ALERT declaration.

C. Incorrect - This condition would result in an ALERT declaration D. Incorrect - 15 minute time is allowed to extinguish the fire prior to making the declaration.

Technical Reference(s): EIP-ZZ-00101, OTO-KC-00001, and EIP-ZZ-00226 References to be provided to applicants during examination:

1. EIP-ZZ-00101 ADD 1, Emergency Action Level Classification Matrix, Rev 3 Learning Objective:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator From T61.0003D, Emergency Operations, LP RERP (SRO-AEO-1 and SRO-RER-2) - Given a set of plant conditions, classify an event using EIP-ZZ-00101 Question Source: Bank # ___L13411___

Modified Bank # ______

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content:

(CFR: 43.5)

Comments: SRO per #5

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Emergency Procedures / Plan Group # Generic K/A # 2.4.37 Importance Rating 4.1 Knowledge of the lines of authority during implementation of the emergency plan.

Question #100 A General Emergency has been declared.

Which of the following activities can be delegated to another individual by the Emergency Coordinator?

A. Classifying and declaring emergencies.

B. Requesting the formation of emergency teams.

C. Authorize personnel exposure limits in excess of 10CFR20 limits.

D. Decision making for implementing strategies identified in the Severe Accident Management Guidelines.

Answer: B Explanation: Per EIP-ZZ-00102, revision 57, step 3.1.2 the EC may delegate the responsibility of initiating implementation of onsite protective actions. Step 3.1.1 lists the responsibilities that may not delegated and is the list of distractors.

A. Incorrect per above explanation B. Correct C. Incorrect per above explanation D. Incorrect per above explanation Technical Reference(s):

1. EIP-ZZ-00102, Emergency Implementing Actions, Rev 57 References to be provided to applicants during examination: None Learning Objective:

1. T61.003D Control Board Certification - Emergency Operations, SD-11 Objective N Perform action of emergency coordinator in the Control Room during an emergency.

2. T68.1020 (6) Emergency Coordinator Control Room (RO/SRO/STA), Objective D Explain the actions to be taken to respond to an emergency classification following an event declaration, per EIP-ZZ-00102 including required paperwork.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Question Source: Bank # ________

Modified Bank # __X L16661____

New _______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _X____

Comprehension or Analysis _____

10 CFR Part 55 Content:

CFR: 43.3 Comments: