ML13160A004

Callaway Plant - 2013-03 - Final Written Exam

ML13160A004
Person / Time
Site:	Callaway
Issue date:	03/26/2013
From:	Vincent Gaddy Operations Branch IV
To:	Union Electric Co
laura hurley
References
Download: ML13160A004 (205)

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NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 008 - Pressurizer Vapor Space Accident Group # 1 K/A # 008 AK3.04 Importance Rating 4.2 Knowledge of the reasons for the following responses as they apply to the Pressurizer Vapor Space Accident: RCP tripping requirements.

Question #1 Given the following plant conditions:

• An RCS small break LOCA has occurred at the weld connection of the Pressurizer Spray line connection at the top of the Pressurizer

• A reactor trip has occurred and the crew is responding in accordance with E-0, Reactor Trip or Safety Injection

Which one of the following describes when and why the Reactor Coolant Pumps (RCPs) would be tripped?

When RCS pressure lowers to less than 1425 psig with-

A. at least one CCP running; to reduce RCS inventory loss.

B. NO ECCS pumps running; to reduce RCS inventory loss.

C. at least one CCP running; to prevent RCP motor damage from containment conditions.

D. NO ECCS pumps running; to prevent RCP motor damage from containment conditions.

Answer: A Explanation:

A. Correct. This is directed by the RCP Trip Criteria found on the Foldout Page of E-0. The basis of this direction is to minimize the loss of RCS inventory during a SBLOCA which is the pressurizer vapor space accident described in the question. B. Incorrect. Reason is correct but ECCS pump status is not correct. With no ECCS pumps running, the RCPs are required to ensure core heat removal.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Incorrect. Incorrect reason given to prevent RCP motor damage. This reason is plausible as the containment will be in a very humid condition but this is not a concern for securing the RCPs. RCS pressure and CCP status is correct. D. Incorrect. Incorrectly states this can be done with no ECCS pumps running. Also has incorrect reason given to prevent RCP motor damage. This reason is plausible as the containment will be in a very humid condition but this is not a concern for securing the RCPs.

Technical Reference(s

): E-0, Reactor Trip or Safety Injection, and ERG Executive Volume-Generic Issue RCP Trip/Restart References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-02, Obj A, Describe the bases for the RCP trip criteria.

Question Source: Bank # ______ Modified Bank # ______ New ____X___

Question History: Last NRC Exam ___________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 009 - Small Break LOCA Group # 1 K/A # 009 EK1.02 Importance Rating 3.5 Knowledge of the operational implications of the following concepts as they apply to the small break LOCA: Use of steam tables.

Question #2 Given the following plant conditions:

• A Reactor Trip and Safety Injection have occurred due to a small break Loss of Coolant Accident

• ES-1.2, Post LOCA Cooldown and Depressurization, is in progress

• Current RCS conditions are as indicated below:

• BB PI-455A, RCS Narrow Range Pressure 1700 psig

• BB PI-456, RCS Narrow Range Pressure 1700 psig

• BB PI-403, RCS Wide Range Pressure 1535 psig

• Highest Core Exit Thermocouple 530°F

• Highest RCS Hot Leg Temperature 510°F Which one of the following choices corre ctly completes the statement below?

Subcooling is monitored on the Fo ldout Page to (1) AND the curr ent value of subcooling is (2) .

A. (1) ensure SI rein itiation, if required (2) 70°F B. (1) prevent voiding in the reacto r vessel head during depressurization (2) 70°F C. (1) ensure SI rein itiation, if required (2) 104°F

D. (1) prevent voiding in the reac tor vessel head during depressurization (2) 104°F NRC Site-Specific Written Examination Callaway Plant Reactor Operator Answer: A Explanation:

A. Correct. Per ES-1.2.Foldout Page criteria, RCS subcooling is monitored for SI reinitiation criteria. Current subcooling for given conditions is: saturated temperature for lowest pressure is 600°F - highest temperature of 530°F equals 70°F. B. Incorrect. Correct subcooling but incorrect reason. Voiding concern is plausible as the procedure caution that it could occur; however there is no subcooling limits given for this concern. C. Incorrect. Reason given is correct but subcooling is incorrect. Subcooling value is plausible if the non-conservative RCS temperature is used in the calculation.

D. Incorrect. Both subcooling and reason are incorrect. See A and B for explanations.

Technical Reference(s

): ES-1.2, Post LOCA Cooldown and Depressurization ERG Executive Volume-Generic Issue Foldout Page Items Steam Tables References to be provided to applicants during examination:

Steam Tables Learning Objective: T61.003D, LP D-10, Obj E, Describe the criteria and the basis for information as stated on the ES-1.2, Post LOCA Cooldown and Depressurization, Foldout Page.

Question Source: Bank # __X____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____Harris 2009 Exam________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content:

55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 011 - Large Break LOCA Group # 1 K/A # 011 EA1.11 Importance Rating 4.2 Ability to operate and monitor the following as they apply to a Large Break LOCA: Long-term cooling of core. Question #3 Given the following plant conditions:

• A large break Loss of Coolant Accident occurred

• 'B' RHR Pump tripped and could not be restarted when ES-1.3, Transfer to Cold Leg Recirculation, was performed

• The crew is in the process of shifting to hot leg recirculation in accordance with ES-1.4, Transfer to Hot Leg Recirculation

Which pump(s) will have to be stopped, then restarted, when performing ES-1.4?

A. 'A' Residual Heat Removal Pump B. 'A' and 'B' Safety Injection Pumps C. 'A' and 'B' Containment Spray Pumps D. 'A' and 'B' Centrifugal Charging Pumps Answer: B Explanation:

A. Incorrect. Plausible as the RHR pumps flowpath is realigned during ES-1.4, but the pumps do not need to be secured.

B. Correct. Flowpath to cold legs are isolated prior to opening hot leg flowpath; pump recirc valves were closed in ES-1.3. C. Incorrect. Only the SI pumps are stopped, then restarted, when performing ES-1.4. D. Incorrect. No changes are made to the CCP flowpath when performing ES-1.4.

Technical Reference(s

): ES-1.4, Transfer to Hot Leg Recirculation

References to be provided to applicants during examination:

None NRC Site-Specific Written Examination Callaway Plant Reactor Operator Learning Objective: T61.003D, LP D-12, Obj G, Describe flowpaths of all the ECCS Pumps when lined up for Hot leg Recirculation.

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: 55.41.8 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 015/017 - Reactor Coolant Pump (RCP) Malfunctions Group # 1 K/A # 015/017 AK1.01 Importance Rating 4.4 Knowledge of the operational implications of the following concepts as they apply to Reactor Coolant Pump Malfunctions (Loss of RC Flow): Natural circulation in a nuclear reactor power plant.

Question #4 The plant is tripped from full power due to a loss of Component Cooling Water. The crew performs all appropriate actions and transitions to ES-0.1, Reactor Trip Response.

The following conditions exists 20 minutes after the trip:

• SG pressures are all approximately 1000 psig and stable

• RCS pressure is 2230 psig and stable

• Thot is approximately 575°F in all loops and slowly lowering

• Core Exit TCs indicate approximately 580°F and stable

• Tcold is approximately 543°F in all loops and stable Which one of the following descr ibes the status of RCS H eat removal for the current plant conditions?

A. Heat removal is via forced circulati on and both the condenser and 10% atmospheric steam dumps

B. Heat removal is via forced circulation and the condenser steam dumps only C. Heat removal is via natural circ ulation and the condenser steam dumps only D. Heat removal is via natural circulation and both the condenser and 10%

atmospheric steam dumps Answer: C Explanation:

A. Incorrect. Conditions are established and indicate that natural circulation has been established. B. Incorrect. RCPs are not running. Loss of CCW for 10 minutes require that the RCPs be secured. C. Correct. Per EOP Addendum 1, Natural Circulation Verification, given conditions indicate that natural circulation has been established.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator D. Incorrect. Heat removal is being maintained by natural circulation and the condenser steam dumps. The setpoint for the ASDs is 1125. The given steam pressure is below the setpoint for the ASDs.

Technical Reference(s

): EOP Addendum 1, Natural Circulation Verification References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP D-2, Obj J, Describe the natural circulation process. Obj K, State the symptoms used to verify natural circulation.

Question Source: Bank # __Diablo Canyon 2007 Exam____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.2 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 025 - Loss of Residual Heat Removal System Group # 1 K/A # 025 2.1.32 Importance Rating 3.8 Ability to explain and apply system limits and precautions.

Question #5 Given the following plant conditions:

• The plant is in Mode 5

• RCS is at reduced inventory for nozzle dam installation with a Hot Core

• RHR Pump 'A', is in service

• At 0318 RHR Pump 'A' tripped due to unknown reasons In accordance with OTO-EJ-00003, Loss of RHR While Operating at Reduced Inventory or Mid-Loop Conditions, which one of the following actions must be taken by the crew?

A. Start PEJ01B, RHR Pump 'B', to provide RCS cooling B. Initiate RWST gravity feed prior to the start of core boiling C. Close the Containment Equi pment Hatch no later than 0418 D. Operate condenser steam dumps to maintain RCS temperature Answer: B Explanation:

A. Incorrect. CAUTION at beginning of OTO-EJ-00003 states not to start a standby RHR pump unless the cause of the loss of flow is known and corrective action has been taken. B. Correct. Action is directed per OTO-EJ-00003, Step 2 RNO.

C. Incorrect. Containment equipment hatch must be closed within 30 minutes. D. Incorrect. Plausible as the use of steam dumps is referenced in loss of RHR procedures, but they would not be an option with nozzle dams installed.

Technical Reference(s

): OTO-EJ-00003, Loss of RHR While Operating at Reduced Inventory or Mid-Loop Conditions

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination:

None Learning Objective:

T61.003E, LP 3, Obj H, State major action categories and symptoms/entry conditions for OTO-EJ-00003, Loss of RHR while operating at reduced inventory or mid loop.

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: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 026 - Loss of Component Cooling Water Group # 1 K/A # 026 AA2.01 Importance Rating 2.9 Ability to determine and interpret the following as they apply to the Loss of Component Cooling Water: Location of a leak in the CCWS.

Question #6 Given the following plant conditions:

• Reactor power is 100%

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

• Annun 77A, REACT DEV, alarms

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

• RCS Tavg is slowly rising

Which one of the following has caused the above indications?

A leak in the-

A. Letdown Heat Exchanger B. 'A' RHR Heat Exchanger C. Seal Water Heat Exchanger D. RCP 'D' Thermal Barrier Heat Exchanger Answer: C Explanation:

A. Incorrect. CCW is the lower pressure in this HEX, thus RCS temp would not be affected. Leakage would be into the CCW system and Annun 51D would be received but for high level. B. Incorrect. Leakage would be into RHR with CCW surge tank level lowering but the RHR system is isolated from the RCS at this power thus there would be no effect on the RCS. C. Correct. Leakage would be from CCW into the CVCS causing a RCS dilution. D. Incorrect. CCW is the lower pressure in this HEX, thus RCS temp would not be affected. Leakage NRC Site-Specific Written Examination Callaway Plant Reactor Operator would be into the CCW system and Annun 51D would be received but for high level.

Technical Reference(s

): OTO-EG-00001, CCW System Malfunction References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 10, Obj H, Describe the plant response and discuss any possible detrimental effects of leaks in the following Heat Exchangers (H/Xs): 1. Letdown H/X 2. Seal Water H/X

3. RHR H/X 4. Fuel Pool Cooling H/X 5. CCW H/X

6. RCP Thermal Barrier H/X 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: 55.41.5 and 7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 027 Pressurizer Pressure Control Malfunction Group # 1 K/A # 027 AK2.03 Importance Rating 2.6 Knowledge of the interrelations between the Pressurizer Pressure Control Malfunctions and the following: Controllers and positioners.

Question #7 Given the following plant conditions:

• Reactor power is 3%

• Pressurizer Pressure Channel, BB PT-457, is controlling Pressurizer pressure

• BB PT-457 fails HIGH Which one of the following descr ibes plant response if NO operator action is taken?

The Reactor will-

A. NOT TRIP due to t he given power level.

B. TRIP on Low Pressurizer Pressure.

C. TRIP on High Pressurizer Pressure.

D. TRIP when Safety Injection actuates.

Answer: D Explanation:

A. Incorrect. Answer is plausible since the given power level will block the low pressure trip but will not block the SI pressure trip at 1849 psig.

B. Incorrect. RCS pressure w ill lower, but the trip is blocked at this power level due to being below the P-7 setpoint of 10%. C. Incorrect. PT-457 failing high causes the pressurizer spray valves to open, thus RCS pressure will lower rather than rise.

D. Correct. The pressurizer spray valves will open due to 457 failing high causing RCS pressure to lower.

The low pressure reactor trip is blocked by P-7. Pressure will lower to the SI setpoint at which time the SI signal will cause a reactor trip.

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

): OTO-SA-00001, Engineered Safety Feature Actuation Verification And Restoration and Drawing 7250D64, Sheet 6, Pressurizer Trip Signals

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 27, Obj C, List all the Reactor Trip Signals supplied to RPS, including setpoint, coincidence, interlocks and protection afforded and Obj D, List all the RPS Permissive Signals, including setpoints, coincidence and function.

Question Source: Bank # __R12060____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.7 Comments: Candidate must understand the interrelationship of a failed pressurizer pressure channel and the Pressurizer Pressure Master Controller, BB PK-455A, and the plant response to the controller output.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 029 - Anticipated Transient Without Scram (ATWS) Group # 1 K/A # 029 EK3.11 Importance Rating 4.2 Knowledge of the reasons for the following responses as they apply to the ATWS: Initiating emergency boration.

Question #8 Which one of the following descr ibes the reason that emergen cy boration is initiated in FR-S.1, Response To Nuclear Power Generation/ATWS?

A. It is performed to minimize void formation in the reactor core.

B. It is performed to add negative reactivity to bring the reactor core subcritical.

C. It is the fastest mechanism for adding negative reactivity to the reactor core.

D. It bypasses the Boric Acid Filter to minimize loss of flow due to filter clogging.

Answer: B Explanation:

A. Incorrect. Plausible as emergency boration is adding a volume of borated water to the RCS that would help fill any voids that might be created due to the failure to trip. Per the background document for FR-S.1, emergency boration is performed to ensure the reactor is subcritical, not to minimize void formation. B. Correct. Per BD-FR-S.1, Response To Nuclear Power Generation/ATWS, the purpose of emergency boration is to add negative reactivity to bring the reactor core subcritical.

C. Incorrect. It is not the fastest method to add negative reactivity; the fastest method is to insert rods. D. Incorrect. Filter clogging would be a concern as it would reduce emergency borate flow, however, any flowpath for emergency boration must go through the BA filter.

Technical Reference(s

): BD-FR-S.1 and FR-S.1, Response To Nuclear Power Generation/ATWS References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-29, Obj A, Explain the Purpose and Major Action Categories of: FR-S.1, Response to Nuclear Power Generation/ATWS

Question Source: Bank # ______ Modified Bank # ______

NRC Site-Specific Written Examination Callaway Plant Reactor Operator New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.5 and 10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 038 - Steam Generator Tube Rupture Group # 1 K/A # 038 2.2.42 Importance Rating 3.9 Ability to recognize system parameters that are entry-level conditions for Technical Specifications.

Question #9 Reactor Power is 80%.

Which one of the following would FAIL to meet the Limiting Condition for Operation (LCO) for RCS Operational Leakage?

A. 0.15 gpm primary to secondary leakage in Steam Generator 'B' B. 0.6361 gpm unidentified leakage listed on t he Shift Manager Operational Focus Items C. 2.4 gpm leakage from RCP 'A' #2 Seal to the Reactor Coolant Drain Tank D. 3.5 gpm leakage from BB PC V-456, PZR PORV, into the Pressurizer Relief Tank

Answer: A Explanation:

A. Correct. The TS limit is 150 gal/day which is equal to .104 gpm. B. Incorrect. Limit for unidentified leakage is 1 gpm.

C. Incorrect. This is identified leakage which has a TS allowance of 10 gpm. D. Incorrect. This is identified leakage which has a TS allowance of 10 gpm.

Technical Reference(s

): Tech Spec 3.4.13, RCS Operational Leakage Tech Spec 3.4.14, RCS Pressure Isolation Valve (PIV) Leakage References to be provided to applicants during examination:

None Learning Objective:

T61.0110, LP 9, Obj P, State the LCOs for the RCS Operational Leakage Technical Specification and Explain the basis (SRO only for basis). Define Identified Leakage, Unidentified Leakage, and Pressure boundary Leakage. Obj Q, State the LCOs for the RCS Pressure Isolation Valve (PIV) Leakage Technical Specifications and Explain the basis (SRO only for basis).

Question Source: Bank # __R12107____

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

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 W/E12 - Uncontrolled Depressurization of all Steam Generators Group # 1 K/A # W/E12 EA1.1 Importance Rating 3.8 Ability to operate and / or monitor the following as they apply to the (Uncontrolled Depressurization of all Steam Generators): Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Question #10 Given the following plant conditions:

• The reactor tripped from 30% power due to a faulted Steam Generator (SG)

• Due to an uncontrolled depressurization of all SGs, the crew implemented ECA-2.1, Uncontrolled Depressuriza tion of All Steam Generators

• Fifteen (15) minutes following the reactor trip the following plant conditions exist:

• RCS Cold Leg Temperature 505°F lowering at a constant rate

• Containment Pressure 20 psig stable

• SG NR Level SG 'A' - 27% SG 'B' - 5%

SG 'C' - 8% SG 'D' - 23%

• The crew is currently at Step 2, Contro l Feed Flow To Mini mize RCS Cooldown

Which SG(s) require(s) a reduction of feed flow to 27,500 lbm/hr?

A. 'A' Only B. 'B' and 'C' Only C. 'A' and 'D' Only D. All Four SGs

Answer: D Explanation:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator A. Incorrect. Due to currently having a cooldown rate >100°F/hr feed flow should be lowered to 27,500 lbm/hr to all 4 SGs; plausible because highest level and >25% NR. B. Incorrect. Due to currently having a cooldown rate >100°F/hr feed flow should be lowered to 27,500 lbm/hr to all 4 SGs; plausible because candidate may think to stop feeding SGs with lowest level. C. Incorrect. Due to currently having a cooldown rate >100°F/hr feed flow should be lowered to 27,500 lbm/hr to all 4 SGs; plausible because these are the SGs with the highest level. D. Correct. All SGs require AFW throttling per ECA-2.1, Step 2 RNO.

Technical Reference(s

): ECA-2.1, Uncontrolled Depressurization Of All Steam Generators

References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-16, Obj F, Explain the Notes and Cautions, including their basis, for ECA-2.1, Uncontrolled Depressurization of all Steam Generators and Obj I, Outline procedural flowpath including major system and equipment operation in accomplishing the goal of ECA-2.1, Uncontrolled Depressurization of all Steam Generators.

Question Source: Bank # __R11890____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 055 - Station Blackout Group # 1 K/A # 055 EA2.06 Importance Rating 3.7 Ability to determine or interpret the following as they apply to a Station Blackout: Faults and lockouts that must be cleared prior to re-energizing buses.

Question #11 A station blackout has occurred.

The crew is attempting to restore power to NB01 via its normal power source, XNB01, in accordance with ECA-0.0, Loss of All AC Power.

Which one of the following relays would have to be MANUALLY reset?

A. Lockout Relay (286-1)

B. Neutral Ground O/C (151N)

C. Differential O/C (287)

D. Switchyard Trip Relay (94-1)

Answer: A Explanation:

A. Correct. This relay must be manually reset. B. Incorrect. This relay will reset automatically. C. Incorrect. This relay will reset automatically.

D. Incorrect. This relay will reset automatically.

Technical Reference(s

): OTA-RK-00016, Addendum 19A, XNB01 Transformer Lockout References to be provided to applicants during examination:

None Learning Objective: T61.GFES, LP-34 (Chap 6), Obj 18, Describe the function of a relay device and Obj 19, Identify conditions for which a relay function is used in circuit breakers.

Question Source: Bank # ______ Modified Bank # ______ New ___X____

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 056 - Loss of Off-Site Power Group # 1 K/A # 056 AA1.21 Importance Rating 3.3 Ability to operate and / or monitor the following as they apply to the Loss of Offsite Power: Reset of the ESF load sequencers.

Question #12 Given the following plant conditions:

• Reactor power is 100%

• A loss of offsite power has occurred

• Both ESF busses, NB01 and NB02, ar e being powered by their respective emergency diesel generators

Which one of the following actions will RESET the Train 'A' Shutdown Sequencer for

NB01?

A. Close NB01 AEPS Supply Breaker NB0114 B. Close NB01 Normal Supply Breaker NB0112 C. Open NB01 Alternate Supply Breaker NB0109 D. Open NB01 Emergency Supply Breaker NB0111

Answer: B Explanation:

A. Incorrect. This breaker can be used to provide a source of power to the bus but the sequencer does not use this breaker for any function. B. Correct. Either the normal or alternate bus feeder breaker must be closed to reset the SD sequencer. C. Incorrect. The open position for this breaker is the normal position for this breaker at power. However, closing this breaker to supply power to the bus will reset the sequencer. D. Incorrect. This breaker must be opened to return the bus to a normal lineup. The closing of this bus was what initiated the SD sequencer but opening it without power to the bus will not reset the sequencer.

Technical Reference(s

): E-22NF01, Load Shedding and Emergency Load Sequencing Logic References to be provided to applicants during examination:

None NRC Site-Specific Written Examination Callaway Plant Reactor Operator Learning Objective: T61.0110, LP 51, Obj B, List the conditions required to actuate AND reset the following: 1. Loss of Coolant Accident (LOCA) Sequencer 2. Shutdown (Blackout) Sequencer.

Question Source: Bank # ______ Modified Bank # ______ New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 057 - Loss of Vital AC Instrument Bus Group # 1 K/A # 057 AK3.01 Importance Rating 4.1 Knowledge of the reasons for the following responses as they apply to the Loss of Vital AC Instrument Bus: Actions contained in EOP for loss of vital ac electrical instrument bus.

Question #13 Given the following plant conditions:

• Reactor power is 100%

• A loss of 120 VAC Instrument Bus NN01 occurs In accordance with OTO-NN-00001, Loss of Safety Related Instrument Power, the operator will be directed to (1) Isolate Letdown and (2) Place Service Air Compressor

'A' in Pull-To-Lock (PTL).

Which one of the following correctly prov ides the reasons for these actions?

(1) Isolate Letdown (2) Air Compressor To PTL A. NCP suction swapped from the VCT to the RWST Loss of cooling water to air compressor B. NCP suction swapped from the VCT to the RWST Loss of control power to air compressor C. Charging Flow Control Valve BG FCV-124 closed Loss of control power to air compressor D. Charging Flow Control Valve BG FCV-124 closed Loss of cooling water to air compressor Answer: A Explanation:

A. Correct. NCP suction has swapped due to BG LT-112, VCT Level Channel, failing low and EF HV-43, ESW Train A To Air Compressor A, has failed closed causing a loss of cooling water to 'A' air compressor. B. Incorrect. Isolating letdown is correct but control power for 'A' air compressor comes from the main power supply feed (NG) via an internal transformer, not NN power.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Incorrect. BG FCV-124 is an air controlled valve thus it is not affected by the loss of NN01; control power for 'A' air compressor comes from the main power supply feed (NG) via an internal transformer, not NN power. D. Incorrect. BG FCV-124 is an air controlled valve thus it is not affected by the loss of NN01. Action for 'A' air compressor is correct.

Technical Reference(s

): OTO-NN-00001, Loss of Safety Related Instrument Power

References to be provided to applicants during examination:

None Learning Objective: T61.003B, LP B-27, Obj C, Given a set of plant conditions or parameters indicating a Loss of Safety Related Instrument Power, Analyze the correct pr ocedure(s) to be utilized and the required actions to stabilize the plant.

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: 55.41.7 and 10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 058 - Loss of DC Power Group # 1 K/A # 058 AK1.01 Importance Rating 2.8 Knowledge of the operational implications of the following concepts as they apply to Loss of DC Power: Battery charger equipment and instrumentation.

Question #14

Given the following plant conditions:

• Reactor power is 100%

• The main feeder breaker to Load Center NG01 trips on overcurrent

• NO operator action has been taken

Which one of the following descr ibes the effect on the uni t electrical alignment?

A. Battery Charger NK25 is powered. 120 VAC Bus NN01 is supplied by Battery NK11.

B. Battery Charger NK21 has lost power.

120 VAC Bus NN01 is supplied by Battery NK11.

C. Battery Charger NK21 is powered.

120 VAC Bus NN01 has transferred to the alternate supply.

D. Battery Charger NK25 has lost power.

120 VAC Bus NN01 has transferred to the alternate supply.

Answer: B Explanation:

A. Incorrect. NK25 is powered from NG01.

B. Correct. NK21 is powered from NG01 and has lost power with battery NK11 now supplying power to NN01 through NK01. C. Incorrect. NK21 has lost power as it is powered by NG01.

D. Incorrect. NN01 has not transferred to alternate power. It is still being supplied by NK01. NK01 is what has lost its normal power supply and is now powered from battery NK11.

Technical Reference(s

): OTN-NK-00001 ADD 01, 125VDC Bus NK01 And Distribution System

References to be provided to applicants during examination:

None NRC Site-Specific Written Examination Callaway Plant Reactor Operator Learning Objective: T61.0110, LP-6, Obj A, Draw and Explain a one line diagram of the Safeguards Power System to include the components and subsystems listed in objective B with normal breaker alignments shown and Obj B, Describe the purpose and operation of the following Safeguards Power system components and subsystems: 5. 125 VDC System (NK)

Question Source: Bank # ___X___ Modified Bank # ______ New _______

Question History: Last NRC Exam ___Wolf Creek 2007 Exam_________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 062 - Loss of Nuclear Service Water Group # 1 K/A # 062 2.4.6 Importance Rating 3.7 Knowledge of EOP mitigation strategies.

Question #15 Given the following plant conditions:

• A Loss of Coolant Accident and Safety Injection have occurred

• The crew has completed E-0, Reacto r Trip or Safety Injection, and have transitioned to E-1, Loss of Reactor or Secondary Coolant

• The crew observes the following condi tions while performing Step 11, Check Ultimate Heat Sink - Normal:

• Annun 30E, NG07 BUS UV/OV, is in alarm

• NG HIS-4, 4.16 KV BUS NB01 To ESW BKR NB0117, has tripped

In accordance with E-1, which one of the following actions must be taken by the Control Room to mitigate this condition?

A. Remove ESW Train 'A' from service

B. Manually Cross-Connect ESW Trains 'A' and 'B'

C. Realign Train 'A' ESW to the Service Water System D. Leave ESW Train 'A' in service whil e monitoring UHS return temperature Answer: A Explanation:

A. Correct. This action is correct in order to maintain operability of ESW Train 'B'. B. Incorrect. Plausible as this can be done, but not procedurally allowed. C. Incorrect. Plausible as this is possible, but would not be done in an emergency situation. This is the normal lineup for SW/ESW.

D. Incorrect. Procedure does not allow for this action. It is plausible as the temperature could be monitored but E-1 directs to secure the affected ESW train for the above conditions.

Technical Reference(s

): E-1, Loss of Reactor or Secondary Coolant

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-8, Obj C, Describe the requirements and basis for the Continuous Action Steps of E-1, Loss of Reactor or Secondary Coolant and Obj F, Explain the Notes and Cautions, including their basis, for E-1, Loss of Reactor or Secondary Coolant.

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: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 W/E11 - Loss of Emergency Coolant Recirculation Group # 1 K/A # W/E11 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 crew has entered ECA-1.1, Loss Of Emergency Coolant Recirculation

• A Containment Spray Actuat ion Signal was received

• Containment Pressure is now 29 psig Which of the following plant conditions wo uld allow BOTH Containment Spray Pumps to be secured?

RWST LEVEL CTMT COOLERS RUNNING IN SLOW SPEED A. 20% 3 B. 20% 2 C. 40% 3 D. 40% 2

Answer: A Explanation:

A. Correct. B. Incorrect. Condition requires 1 CS pump to be in service IAW ECA-1.1.

C. Incorrect. Condition requires 1 CS pump to be in service IAW ECA-1.1. D. Incorrect. Condition requires 1 CS pump to be in service IAW ECA-1.1.

Technical Reference(s

): ECA-1.1, Loss of Emergency Coolant Recirculation

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination:

Learning Objective: T61.003D, LP D-13, Obj H, Outline procedural flowpath including major system and equipment operation in accomplishing the goal of ECA-1.1, Loss of Emergency Coolant Recirculaiton.

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: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 W/E05 - Loss of Secondary Heat Sink Group # 1 K/A # W/E05 EK2.2 Importance Rating 3.9 Knowledge of the interrelations between the (Loss of Secondary Heat Sink) and the following: Facility's heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operation of the facility.

Question #17 Given the following plant conditions:

• A Reactor Trip and Safety Injection occurred from 100%

• The crew has entered FR-H.1, Response To Loss Of Secondary Heat Sink

• Current plant conditions are:

• RCS pressure 1600 psig and stable

• Steam Generator (SG) pressures 1040 psig and stable in all SGs

• SG 'A', 'B' and 'D' Wide Range level 25% and slowly lowering

• SG 'C' Wide Range level 30% and slowly lowering

• Total Auxiliary Feedwater (AFW) flow 0 gpm

In accordance with FR-H.1, which one of the following actions is required to be taken by

the crew?

A. Initiate secondary depressurization to establish condensate flow.

B. Trip reactor coolant pumps and initiate RCS Bleed and Feed.

C. Perform EOP Addendum 38, Non Safety Auxiliary Feedwater Pump, to establish AFW flow.

D. Continue efforts to establish AFW flow while monitoring SG water levels for RCS Bleed and Feed criteria.

Answer: B Explanation:

A. Incorrect. Plausible as this would be action directed in FR-H.1 if bleed and feed criteria was not met. B. Correct. Action is correct IAW Step 2 of FR-H.1.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Incorrect. Plausible as this would be the preferred action in FR-H.1 if bleed and feed criteria was not met. D. Incorrect. Plausible as this would be the correct action if Bleed and Feed criteria are not met. Having 3 of 4 SGs below 27% WR level satisfies Bleed and Feed criteria. All SGs at this level are not required.

Technical Reference(s

): FR-H.1, Response To Loss Of Secondary Heat Sink References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-26, Obj D, Describe the requirements and basis for the continuous action steps of FR-H.1, Response To Loss Of Secondary Heat Sink.

Question Source: Bank # __L14388____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.10 Comments: Two distractors modified for better plausibility.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 077 - Generator Voltage and Electric Grid Disturbances Group # 1 K/A # 077 AA2.10 Importance Rating 3.6 Ability to determine and interpret the following as they apply to Generator Voltage and Electric Grid Disturbances: Generator overheating and the required actions.

Question #18

Given the following plant condition s following a grid disturbance:

• Main Generator H 2 Pressure 75 psig

• Voltage Regulator AUTOMATIC

• Main Generator Megawatts (MWs) 1100 MWe

• Megavars (MVARs) 800 MVARs

Which one of the following is the:

(1) Equipment concern associated with the above conditions?

AND

(2) Required action to be taken by t he Balance of Plant (BOP) Operator?

(Refer to Curve Book Figures 10-1 thru 4, which are provided for reference)

(1) Equipment Concern (2)

BOP Action A. Generator Armature Heating REDUCE Main Generator MWs to 950 B. Generator Field Heating REDUCE Main G enerator MVARs to 650 C. Generator Armature Heating REDUCE Main Generator MVARs to 650 D. Generator Field Heating REDUCE Main Generator MWs to 950

Answer: B Explanation:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator A. Incorrect. Action is correct but concern is incorrect. B. Correct. Required action and equipment concern are correct IAW Figure 10-2. C. Incorrect. Action would be correct if wrong figure is used (75 lb H2 pressure curve rather than 60 lb curve). Concern is not correct.

D. Incorrect. Answer would be correct if wrong figure is used (75 lb H2 pressure curve rather than 60 lb curve).

Technical Reference(s

): Curve Book Figure 10-1 thru 4, Minimum Excitation Limits at Indicated Terminal Voltages References to be provided to applicants during examination: Curve Book Figure 10-1 thru 4, Minimum Excitation Limits at Indicated Terminal Voltages Learning Objective: T61.GFES, LP-33, Obj 20, Describe the effect of changing the excitation of a generator connected to the grid. T61.0110, LP-34, Obj G, Explain the Generator Operating Limitations associated with OTN-AC-00001, Main Turbine and Generator Systems.

Question Source: Bank # __ R11829____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.10 Comments: Question modified to include grid disturbance in the stem and asked about what the limiting factor is for the maximum loading. Distractors modified to include new values for plausibility and to provide the answer and distractors for the limiting component.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 001 - Continuous Rod Withdrawal Group # 2 K/A # 001 AK1.21 Importance Rating 2.9 Knowledge of the operational implications of the following concepts as they apply to Continuous Rod Withdrawal: Integral rod worth.

Question #19

Given the following plant conditions:

• Reactor power is 75%

• Control Rods are in Automatic

• Control Bank 'D' Rods are at 115 Steps

• Axial Flux Difference (AFD) is 0.0 Which one of the following would result in th e largest INITIAL rise in BOTH Reactor power and Pressurizer level?

A. A continuous rod withdraw al at End of Life (EOL)

B. A S/G safety valve fails OPEN at End of Life (EOL)

C. A S/G safety valve fails OPEN at Beginning of Life (BOL)

D. A continuous rod withdrawal at Beginning of Life (BOL)

Answer: A Explanation:

A. Correct. Results in largest positive reactivity input to the RCS due to magnitude of integral rod worth being larger at EOL versus BOL. Positive reactivity input results in power rising and pressurizer level rising. B. Incorrect. Would result in power rising but initial pressurizer level would lower until rods started to move out to raise temperature from the RCS cooldown.

C. Incorrect. Same reason as Option B. Greater plant impact at EOL for a steam break but pressurizer level will initially lower whether at BOL or EOL. D. Incorrect. Integral rod worth is less at BOL, therefore largest impact will be at EOL, or Option A.

Technical Reference(s

): Curve Book Table 2-2, Callaway Cycle 19 Redesign Summary of Control Rod Worths (pcm)

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination:

None Learning Objective: T61.GFES, LP-41 (Chap 5), Obj 9, Explain the direction of change in the magnitude of control rod worth (CRW) for a change in moderator temperature, boron concentration, and fission product poisons.

Question Source: Bank # ______ Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.1 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 024 - Emergency Boration Group # 2 K/A # 024 AA1.26 Importance Rating 3.3 Ability to operate and / or monitor the following as they apply to Emergency Boration: Boric acid storage tank.

Question #20

Given the following plant conditions:

• A manual Reactor trip was initiated from 100% power

• Three control rods did not insert following the Reactor trip

• Boric Acid Tank 'A' is drained for maintenance

• Boric Acid Tank 'B' contains 19,158 gallons

• Refueling Water Storage Tank (RWST) contains 400,000 gallons

It is estimated to require either 7000 gallons from the Boric Acid Storage System or 30,000 gallons from the RWST.

Which one of the followed is the PREF ERRED action to respond to the above conditions and how long will this action take to enter a Tech Spec?

Emergency borate the RCS from the-A. Boric Acid Storage Tanks at 50 gpm. Tech Spec LCO entered in 1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />.

B. RWST at 100 gpm. Tech Spec LCO entered in 30 minutes.

C. Boric Acid Storage Tanks at 50 gpm. Tech Spec LCO entered in 30 minutes.

D. RWST at 100 gpm. Tech Spec LCO entered in 1.0 hour0 days <br />0 hours <br />0 weeks <br />0 months <br />.

` Answer: C Explanation:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator A. Incorrect. Boration source is correct but time to reach TS/FSAR limit is for the RWST. B. Incorrect. Source for boration is incorrect and time is for the BATS and not the RWST. C. Correct. BA system minimum level per FSAR is 17,658 gal. 19158-17658=1500; at 50 gpm the FSAR limit will be reached in 30 min. Also, per ES-0.1, emergency boration from t he BATS is the preferred method of boration. D. Incorrect. The time is correct for the RWST but this is not the preferred method for emergency boration.

Technical Reference(s

): ES-0.1, Reactor Trip Response EOP Addendum 04, Emergency Boration FSAR 16.1.2.6, Borated Water Sources - Operating Limiting Condition for Operation References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-6, Obj D, Identify and Describe Addenda associated with ES-0.1, Reactor Trip Response T.0110, LP 11, Obj T, List the Limiting Conditions for Operation (LCO) and Explain the bases (SRO only for bases) associated with Boraton Flow Paths and BAST from the FSAR.

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:

55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 036 - Fuel Handling Incidents Group # 2 K/A # 036 AA2.02 Importance Rating 3.4 Ability to determine and interpret the following as they apply to the Fuel Handling Incidents: Occurrence of a fuel handling incident.

Question #21 Given the following plant conditions:

• The plant is in Mode 6

• The reactor core is being o ffloaded to the Fuel Building

• GG RE-27 and 28, Aux/Fuel Building Radiation Monitors, are rising

Which one of the following actuat ion signals could be expected?

A. Safety Injection Signal (SIS)

B. Containment Isolation Signal-A (CISA)

C. Containment Purge Is olation Signal (CPIS)

D. Control Room Ventilati on Isolation Signal (CRVIS)

Answer: D Explanation:

A. Incorrect. A Safety Injection would not be expected for a Fuel Handling incident. B. Incorrect. CISA does not have any input on radiation levels. C. Incorrect. Would not be expected on a Fuel Handling incident in the Fuel Building: would be expected if the FA was dropped and damaged in the Containment. D. Correct. Rising radiation levels on 27 and 28 will actuate a FBIS which will actuate a CRVIS.

Technical Reference(s

): OTO-SA-00001, Engineered Safety Feature Actuation Verification And Restoration

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 60, Obj B, List the conditions that actuate the CRVIS.

Question Source: Bank # ______ Modified Bank # ______

NRC Site-Specific Written Examination Callaway Plant Reactor Operator New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 037 - Steam Generator Tube Leak Group # 2 K/A # 037 AA2.16 Importance Rating 4.1 Ability to determine and interpret the following as they apply to the Steam Generator Tube Leak: Pressure at which to maintain RCS during S/G cooldown.

Question #22

Given the following plant conditions:

• A primary to secondary leak has occurred

• The plant has been shut down and the cr ew is currently performing ES-0.1, Reactor Trip Response, and OTO-BB-00001, Steam Generator Tube Leak, in parallel In accordance with OTO-BB-00001, wh ich of the following is the:

(1) target for RCS pressure when depressurizing the RCS?

AND

(2) reason for this target?

A. (1) RCS pressure should not be lower ed to less than the affected SG pressure (2) Ensure the RCS does not reach sa turated condition during the cooldown B. (1) Depressurize the RCS until RCS subcooling is between 80°F - 100°F (2) Ensure the RCS does not reach sa turated condition during the cooldown C. (1) Depressurize the RCS until RCS subcooling is between 80°F - 100°F (2) To prevent initiation of backfill earlier than anticipated D. (1) RCS pressure should not be lower ed to less than the affected SG pressure (2) To prevent initiation of backfill earlier than anticipated

Answer: D Explanation:

A. Incorrect. Pressure target is correct but wrong reason. Saturation will be based on RCS conditions, not SG pressure.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Incorrect. Target pressure should be lowered to as close to SG pressure as possible without going below SG pressure and maintaining RCS subcooling between 30 deg F and 40 deg F. C. Incorrect. Target pressure should be lowered to as close to SG pressure as possible without going below SG pressure and maintaining RCS subcooling between 30 deg F and 40 deg F.

D. Correct. Caution prior to depressurizing the RCS at Step 15 in OTO-BB-00001 states that RCS pressure should not be lowered to less than affected SG pressure.

Technical Reference(s

): OTO-BB-00001, Steam Generator Tube Leak BD-OTO-BB-00001, Steam Generator Tube Leak References to be provided to applicants during examination:

None Learning Objective: T61.003B, LP B-10, Obj E, Discuss the cautions and notes contained in OTO-BB-00001, Steam Generator Tube Leak.

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: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 067 - Plant Fire on Site Group # 2 K/A # 067 2.4.46 Importance Rating 4.2 Ability to verify that the alarms are consistent with the plant conditions.

Question #23

The following alarm is received on Fire Protection Panel KC-008:

• 4:ZN28A ELEC FIRE PMP 1 RUNNING NO operator actions have been taken in response to the alarm.

A fire in which one of the following pl ant locations would cause this alarm?

A. Reactor Building B. Auxiliary Boiler Room C. NB01 ESF Switchgear Room D. Turbine Driven AFW Pump Room Answer: B Explanation:

A. Incorrect. Reactor Building fire supply valve is a normally closed valve which must be opened by the Control Room operator to initiate system flow. No fire pump start signals will be received by the fire pumps until flow is initiated to the RB by the Control Room and system pressure lowers to the low pressure fire pump start setpoint.

B. Correct. Room is covered by an automatic wet pipe sprinkler system which would automatically actuate on a fire, which would start the electric fire pump. C. Incorrect. Room is protected by halon, not water.

D. Incorrect. Room is protected by a manually actuated system which has a normally closed valve which must be opened locally by an operator. No fire pump start signals will be received by the fire pumps until flow is initiated to the AFW pump room by the OT and system pressure lowers to the low pressure fire pump start setpoint.

Technical Reference(s

): OTA-KC-01008, Addendum 6, KC-008 Message File References to be provided to applicants during examination:

None NRC Site-Specific Written Examination Callaway Plant Reactor Operator Learning Objective: T61.0110, LP 35, Obj G, List the types of Sprinkler Systems used at Callaway and Explain the purpose and operation of each type.

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: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 068 - Control Room Evacuation Group # 2 K/A # 068 AK3.12 Importance Rating 4.1 Knowledge of the reasons for the following responses as they apply to the Control Room Evacuation: Required sequence of actions for emergency evacuation of control room.

Question #24

The Control Room has been evacuated due to a fire in Panel RL005.

The crew is in the process of transferring pl ant control from the Control Room to the Auxiliary Shutdown Panel in accordance with OTO-ZZ-00001, Control Room Inaccessibility.

Which one of the following correctly descri bes equipment operation as directed in OTO-ZZ-00001?

A. An Essential Service Water Pump must be in service prior to starting an Auxiliary Feedwater Pump to ensure an alternate suction supply is available.

B. A Component Cooling Water Pump must be in service prior to starting a Centrifugal Charging Pump (CCP) to suppor t CCP oil cool er operation.

C. An Essential Service Water Pump must be in service prior to starting an Emergency Diesel Generator (EDG) to ens ure EDG lube oil cooling.

D. A Component Cooling Water Pump must be in service prior to restarting a Reactor Coolant Pump (RCP) to support RCP motor cooling.

Answer: B Explanation:

A. Incorrect. ESW is the alternate suction supply for AFW but is started automatically if the conditions require it and the applicable start logic is met.

B. Correct. A note in OTO-ZZ-00001 alerts the operator that a CCP should not be started until CCW is in service to support CCP oil cooler operation. C. Incorrect. It is true that ESW is required for EDG operation, but OTO-ZZ-00001 starts the EDG prior to ensuring that an ESW pump is in operation. D. Incorrect. The RCPs are secured in OTO-ZZ-00001 and the procedure never addresses restarting them.

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

): OTO-ZZ-00001, Control Room Inaccessibility References to be provided to applicants during examination:

None Learning Objective: T61.003B, LP B-31, Obj F, Discuss the major cautions and notes contained in the body and attachments of OTO-ZZ-00001, Control Room Inaccessibility.

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: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 W/E14 - High Containment Pressure Group # 2 K/A # W/E14 EK2.1 Importance Rating 3.4 Knowledge of the interrelations between the (High Containment Pressure) and the following: Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Question #25

Given the following plant conditions:

• Reactor power is 100%

• A Loss of Coolant Accident occurs

• Containment pressure rises to 20 psig The Containment Spray Pumps _____(1)______ start.

A Steam Line Isolation Signal (SLIS) _____(2)______ occur.

(1)

(2) A. WILL NOT WILL B. WILL WILL C. WILL NOT WILL NOT D. WILL WILL NOT Answer: A Explanation:

A. Correct. CS pumps will not start as a CSAS actuates at 27 psig. A SLIS is actuated at 17 psig. B. Incorrect. Containment pressure of 20 psig will not initiate a Containment Spray Actuation Signal - CS pumps will not start.

C. Incorrect. A SLIS will occur at 17 psig. D. Incorrect. Incorrect for spray pumps; incorrect for SLIS.

Technical Reference(s

): OTO-SA-00001, Engineered Safety Feature Actuation Verification And Restoration 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 G, Describe those conditions that will generate a containment spray actuation signal (CSAS) and Identify the Containment Spray system response to a CSAS. T61.0110, LP-20, Obj F, Recognize the plant conditions that result in a Steam Line Isolation Signal (SLIS). 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: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator

+Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 W/E01 - Rediagnosis Group # 2 K/A # W/E01 EA1.1 Importance Rating 3.7 Ability to operate and / or monitor the following as they apply to the (Reactor Trip or Safety Injection/Rediagnosis): Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Question #26

Which of the following plant parameters would be monitored in ES-0.0, Rediagnosis, by the operator to transition to the appropriate post accide nt recovery procedure?

A. Pressurizer Level B. Pressurizer Pressure C. Containment Pressure D. Steam Generator Pressure

Answer: D Explanation:

A. Incorrect. PZR level is not monitored. E-1 transition is based on not having indications to transition to an E-2 or E-3 series procedure. B. Incorrect. PZR level is not monitored. E-1 transition is based on not having indications to transition to an E-2 or E-3 series procedure.

C. Incorrect. Containment pressure is not monitored for any transition from ES-0.0. D. Correct. SG pressure is monitored in ES-0.0 by the operator in order to transition to the appropriate post accident recovery procedure.

Technical Reference(s

): ES-0.0, Rediagnosis References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-5, Obj D, State and Explain the parameters which are evaluated, including their Criteria and Basis, to transition from ES-0.0 to other procedures.

Question Source: Bank # ______ Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____________

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 1 W/E09 - Natural Circulation Operations Group # 2 K/A # W/E09 EK3.3 Importance Rating 3.5 Knowledge of the reasons for the following responses as they apply to the (Natural Circulation Operations): Manipulations of controls required to obtain desired operating results during abnormal, and emergency situations.

Question #27

Given the following plant conditions:

• A loss of offsite power has occurred from 100% power

• E-0, Reactor Trip or Safety Injecti on, and ES-0.1, Reactor Trip Response, have been completed

• A plant cooldown has been initiated usi ng ES-0.2, Natural Circulation Cooldown

• All Reactor Coolant loops ar e available for the cooldown

(1) In accordance with ES-0.2, the ma ximum cooldown rate allowed is -

(2) The basis for this limit is to -

A. (1) 50°F/hr. (2) avoid the possibility of upper head void formation.

B. (1) 50°F/hr.

(2) provide a margin to brittle failure of the reactor vessel.

C. (1) 100°F/hr. (2) avoid the possibility of upper head void formation.

D. (1) 100°F/hr.

(2) provide a margin to brittle failure of the reactor vessel.

Answer: A Explanation:

A. Correct. ES-0.2 limits the CD to 50°F/hr with background reason given to avoid the possibility of upper head void formation. B. Incorrect. CD rate is correct but the basis is for the TS cooldown limit C. Incorrect. CD rate is the TS allowed CD rate and the basis is correct for the 50° limit in ES-0.2. D. Incorrect. CD rate and basis are correct for the TS limit, not the limit in ES-0.2.

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

): ES-0.2, Natural Circulation Cooldown BD-ES-0.2, Natural Circulation Cooldown

References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-7, Obj H, Outline procedural flow path including major system and equipment operation in accomplishing the goal of the following procedures: 1. ES-0.2.

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: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 003 - Reactor Coolant Pump System Group # 1 K/A # 003 K3.03 Importance Rating 2.8 Knowledge of the effect that a loss or malfunction of the RCPS will have on the following: Feedwater and emergency feedwater.

Question #28

The plant was operating at 25% when 'D' RCP developed hi gh vibration and continues to slowly rise.

Which one of the following st atements identifies the operator response necessary to control level in 'D' Steam Generator (SG) when the 'D' RCP is secured?

After the transient, feedwater flow to 'D' SG will be controlled by the- A. Main Feedwater Reg Valve and will be GREATER THAN the flow to the other 3 SGs.

B. Main Feedwater Reg Valve and will be LESS THAN the flow to the other 3 SGs.

C. Main Feedwater Bypass Reg Valve and will be GREATER THAN the flow to the other 3 SGs.

D. Main Feedwater Bypass Reg Valve and will be LESS THAN the flow to the other 3 SGs. Answer: D Explanation:

A. Incorrect. MFW flow will be controlled by the MFW byp valve at this power and the flow will lower as steam flow lowers in the affected SG.

B. Incorrect. MFW flow will be controlled by the MFW byp valve at this power.

C. Incorrect. Flow will lower as steam flow lowers in the affected SG. D. Correct. MFW flow will be controll ed by the MFW byp valve at this power and the flow will lower as steam flow lowers in the affected SG.

Technical Reference(s

): OTN-BB-00003, Reactor Coolant Pumps

References to be provided to applicants during examination:

None NRC Site-Specific Written Examination Callaway Plant Reactor Operator Learning Objective: T61.003B, LP SB-13, Obj C, Respond to off-normal RCP conditions.

Question Source: Bank # __X____ Modified Bank # ______

New _______

Question History: Last NRC Exam ___Watts Bar 2004 Exam_________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.2 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 004 - Chemical and Volume Control System Group # 1 K/A # 004 K6.17 Importance Rating 4.4 Knowledge of the effect of a loss or malfunction on the following CVCS components: Flow paths for emergency boration.

Question #29

Given the following plant conditions:

• Reactor power is 100%

• A transient occurs requiring Emergency Bo ration 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

Which one 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. BG V177, Alternate Immediate Boration Valve C. BN LCV-112D, Charging Pump Suction From RWST D. BG HV-8104, Emergency Borate To Charging Pump Suction Answer: C Explanation:

A. Incorrect. Plausible if candidate doesn't know that the boric acid filter is in series with this valve. B. Incorrect. Plausible if candidate doesn't know that the boric acid filter is in series with this valve. C. Correct. Only available flowpath with a clogged boric acid filter.

D. Incorrect. Plausible if candidate doesn't know that the boric acid filter is in series with this valve.

Technical Reference(s

): M-22BG03, Chemical and Volume Control System M-22BG05, Chemical and Volume Control System

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 11, Obj B, Describe the purpose, operation and interlocks for the following CVCS components: 15. Charging Pump Suction Valves NRC Site-Specific Written Examination Callaway Plant Reactor Operator Obj 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 Obj J, Describe the sequence of events that occur in the RMW System upon initiation and completion of each of the following operational modes. 6. Immediate Boration 7. Alternate Immediate Boration.

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: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 004 - Chemical and Volume Control System Group # 1 K/A # 004 A1.04 Importance Rating 3.9 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CVCS controls including: PZR pressure and level.

Question #30

Given the following plant conditions:

• Reactor power is 100%

• Pressurizer level is LOWERING

• VCT level is RISING

• RCS temperature and pressure are STABLE (1) The event in progress is a-

AND

(2) The long term plant response if NO oper ator action is taken will be to trip on-A. (1) charging line break lo cated at the NCP discharge.

(2) low pressurizer pressure.

B. (1) charging flow control valve failed closed.

(2) low pressurizer pressure.

C. (1) charging flow control valve failed closed.

(2) high pressurizer level.

D. (1) charging line break lo cated at the NCP discharge.

(2) high pressurizer level.

Answer: C Explanation:

A. Incorrect. VCT level would not be trending up. Pressurizer level would lower until letdown isolated and the heaters turned off at which time pressure would lower until the trip setpoint.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Incorrect. Pressurizer level will lo wer until letdown isolates at 17%. Pressurizer level will then rise until it reaches the high level trip setpoint of 92%. The B/U heaters will reenergize when level rises above 17%, which prevents pressure from lowering. VCT level would be trending up. C. Correct. Pressurizer level will lower until letdown isolates at 17%. Pressurizer level will then rise until it reaches the high level trip setpoint of 92%. The B/U heaters will reenergize when level rises above 17%, which prevents pressure from lowering. D. Incorrect. VCT level would not be trending up. Pressurizer level would lower until letdown isolated. Pressurizer level will then rise until it re aches the high level trip setpoint of 92%.

Technical Reference(s

):

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 11, Obj B, Describe the purpose, operation and interlocks for the following CVCS components: 18, CCP Flow Control Valve (FCV), and Obj 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.

Question Source: Bank # __L14386____ Modified Bank # ______

New _______

Question History: Last NRC Exam __Callaway 2005__________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.7 Comments: Second part of question modified to include expected reactor trip to satisfy K/A description (to prevent exceeding design limits).

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 005 - Residual Heat Removal System Group # 1 K/A # 005 K2.03 Importance Rating 2.7 Knowledge of bus power supplies to the following: RCS pressure boundary motor-operated valves.

Question #31

Which one of the following is the power supply to BB PV-8702B, RHR Suction Valve From Loop 4 Hot Leg?

A. NG01B B. NG02B C. NG03C D. NG04C Answer: B Explanation:

A. Incorrect. Plausible as this is the power supply for the two RHR iso valves that are in series with the two RCS valves. B. Correct. Power supply for 8702B is NG02B.

C. Incorrect. Plausible as it is a safety related 480v AC MCC. D. Incorrect. Plausible as it is a safety related 480v AC MCC.

Technical Reference(s

): E-23BB12B, RHR Loop 2 Inlet Isolation Valve

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-7, Obj B, Describe the purpose and operation of the following RHR System components, to include interlocks, controller operation and power supplies. 3. Reactor Coolant System (RCS) Hot Leg Suction Valves to RHR.

Question Source: Bank # __X____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____Wolf Creek 2011 Exam________

Question Cognitive Level:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 006 - Emergency Core Cooling System Group # 1 K/A # 006 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:

• A Loss of Coolant Accident has occurred

• Safety Injection initiated automatically

• RWST level is 48% and LOWERING

• Reactor Coolant System pre ssure is 550 psig and STABLE Which one of the following correct ly states the sources of E CCS flow into the Reactor Coolant System Cold Legs?

A. SI Pumps only B. SI and RHR Pumps only C. CCPs and SI Pumps only D. CCPs, SI Pumps and RHR Pumps Answer: C Explanation:

A. Incorrect. CCPs would also be injecting at this pressure. B. Incorrect. RHR will not be injecting at this pressure. RHR shutoff head is 195 psig. C. Correct. Only SI pumps and CCPs injecting at this pressure. D. Incorrect. RHR will not be injecting at this pressure. RHR shutoff head is 195 psig.

Technical Reference(s

): OTN-EJ-00001, Residual Heat Removal System References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-56, Obj G, Draw and/or Label the normal ECCS lineup and flowpath for Cold Leg Injection.

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

Question History: Last NRC Exam ___Callaway 2011 Exam (Modified)_________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content

55.41.8 Comments: Stem modified to be in ECCS injection mode rather than Hot Leg Recirculation. RCS pressure also changed to be higher, which results in a different correct answer.

Original Question:

Given the following:

• A Large Break LOCA has occurred

• Safety Injection was initiated Automatically

• ECCS is in Hot Leg Recirculation lineup

• Reactor Coolant System Pre ssure is currently 150 psig

Which ONE (1) of the following correctly states the sources of ECCS flow into the Reactor Coolant System Hot Legs?

A. SI Pumps only

B. CCPs and SI Pumps only

C. CCPs, SI Pumps, and RHR Pumps only D. SI Pumps and RHR Pumps only Answer: D Explanation:

When the ECCS is realigned fo r Hot Leg Recirc, RHR Pumps will discharge to the Hot Legs directly AND to the SI Pump / CCP suctions. SI Pumps will discharge to the Hot Legs, but CCPs will still discharge to the Cold Legs.

A. Incorrect. RCS pressure is less that RHR shutoff head, so RHR pumps will also inject flow directly. B. Incorrect. CCPs do not discharge to the Hot Legs C. Incorrect. CCPs do not discharge to the Hot Legs D. Correct.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Technical

References:

LP EJ (RHR); LP EJ (ECCS)

References to be provided to applicants during exam: None Learning Objective: EJ.I Question Source: Bank # (note changes; attach parent) Modified Bank # New Question History:

Last NRC Exam Question Cognitive Level: Memory/Fundamental Comprehensive/Analysis 10CFR Part 55 Content: 55.41.8 NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 007 - Pressurizer Relief Tank / Quench Tank System Group # 1 K/A # 007 A4.10 Importance Rating 3.6 Ability to manually operate and/or monitor in the control room: Recognition of leaking PORV/code safety.

Question #33

Given the following plant conditions:

• Reactor power is 100%

• Pressurizer Relief Tank level 88% and RISING SLOWLY

• Pressurizer Relief Tank pressure 20 PSIG and RISING SLOWLY

• Pressurizer Relief Tank temperature 175°F and RISING SLOWLY Which one of the following is the cause of the conditions above?

A. Pressurizer safety (BB-8010C) leaking by B. RHR discharge relief (EJ-8856A) leaking by C. Closure of Seal Return Isolation (BG HV-8112)

D. Opening of PRT Reactor Mak eup Water Supply (BB HV-8045)

Answer: A Explanation:

A. Correct. A leaking PZR safety would cause PRT level, temp and pressure to rise. B. Incorrect. RHR relief valve not connected to RCS in Mode 1.

C. Incorrect. PRT temp would not be rising. Seal leakoff would be going into the PRT and that would be at the temp of seal injection which has a high temp alarm of 135°F. D. Incorrect. This water would be cold; PRT temp would not be rising.

Technical Reference(s

): M-22BB02, Reactor Coolant System References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-9, Obj C, Draw, Label and Explain a one line diagram of the RCS to include the components listed in objective B showing system alignment under the following conditions: 1. Normal operations, and Obj E, List the inputs to the PRT.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Source: Bank # __0110090E02A (L4559)____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.3 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 008 - Component Cooling Water System Group # 1 K/A # 008 K1.01 Importance Rating 3.1 Knowledge of the physical connections and/or cause-effect relationships between the CCWS and the following systems: SWS.

Question #34

Component Cooling Water 'B' HX is aligned as follows:

• EFHV0052, ESW TRN B TO CCW HX B HV, is Open

• EFHV0060, ESW TRN B FROM CCW HX B HV, is Closed Which one of the following a lignments is required for continuous operation of the Component Cooling Water System when Serv ice Water is supplying the Essential Service Water system?

Number of Service Water Pumps Running 'A' CCW HX ESW Valves A. 1 EFHV0051, ESW TRN A TO CCW HX A HV, is Open EFHV0059, ESW TRN A FROM CCW HX A HV, is Closed

B. 1 EFHV0051, ESW TRN A TO CCW HX A HV, is Open EFHV0059, ESW TRN A FROM CCW HX A HV, is Open

C. 2 EFHV0051, ESW TRN A TO CCW HX A HV, is Open EFHV0059, ESW TRN A FROM CCW HX A HV, is Open

D. 2 EFHV0051, ESW TRN A TO CCW HX A HV, is Open EFHV0059, ESW TRN A FROM CCW HX A HV, is Closed Answer: D Explanation:

A. Incorrect. Not allowed as 1 service water pump cannot supply ESW system loads.

B. Incorrect. Not allowed as 1 service water pump cannot supply ESW system loads. C. Incorrect. With both CCW HEXs in service, both CCW HEX outlet valves must be closed to support service water system operation.

D. Correct. This lineup is allowed by OTN.

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

): OTN-EA-00001, Service Water System

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-4, Obj E, Explain the precautions and limitations associated with the Circulating and Service Water Systems.

Question Source: Bank # __R12310____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 008 - Component Cooling Water System Group # 1 K/A # 008 A3.02 Importance Rating 3.2 Ability to monitor automatic operation of the CCWS, including: Operation of the CCW pumps, including interlocks and the CCW booster pump.

Question #35

Given the following plant conditions:

• Reactor power is 100%

• Component Cooling Water Pump 'D' is Running

• Component Cooling Water Pump 'B' is in Standby

• A Lockout occurs on the Startup Transformer Which one of the following de scribes the design response of the Component Cooling Water (CCW) System?

'D' CCW Pump (1) . 'B' CCW Pump (2) .

(1) (2) A. is shed. remains in standby.

B. continues to run. remains in standby.

C. is shed. is started by the Shutdown Sequencer.

D. continues to run. is st arted by the Shutdown Sequencer.

Answer: C Explanation:

A. Incorrect. 'B' pump will be started by the shutdown sequencer. B. Incorrect. A load shed will occur due to the loss of NB02. C. Correct. NB02 is fed by the SUT. A load shed will occur, 'B' CCW pump will start and the 'D' pump will not restart.

D. Incorrect. A load shed will occur due to the loss of NB02.

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

): E-22NF01, Load Shedding and Emergency Load Sequencing Logic References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-10, Obj B, Draw and/or Label a one line diagram of the CCW System to include the components listed in objective C showing the system alignment under the following conditions: 1. Normal operations.

Question Source: Bank # __0110100C10A (L4619)____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 010 - Pressurizer Pressure Control System Group # 1 K/A # 010 K6.01 Importance Rating 2.7 Knowledge of the effect of a loss or malfunction of the following will have on the PZR PCS: Pressure detection systems.

Question #36

Given the following plant conditions:

• Reactor power is 100%

• BB PT-455, PZR Pressure Transmi tter, had previously failed HIGH

• All actions in OTO-BB-00006, Pressurizer Pressure Control Malfunction, for BB PT-455 are complete EXCEPT the Bistables have NOT been tripped

• Pressure control is selected for BB PT-456 & 457

• Pressurizer spray valves are in MANUAL

BB PT-456, PZR Pressure Transmitter, fails HIGH.

Which one of the following de scribes the response of BB PCV-455A, Pressurizer PORV, and the plant when BB PT-456 fails HIGH?

(Assume NO operator action.)

PORV 455A Plant Response A. Opens Reactor Does Not Trip B. Opens Reactor Trip C. Remains closed Reactor Trip D. Remains closed Reactor Does Not Trip Answer: B Explanation:

A. Incorrect. PORV does open because 455 high trips one of the B/Ss and the controlling channel 456 trips the remaining B/S. Reactor does trip due to two channels indicating high RCS pressure. B. Correct. PORV does open because 455 high trips one of the B/Ss and the controlling channel 456 trips the remaining B/S. 2/4 high pressure B/S logic satisfied resulting in a reactor trip on high NRC Site-Specific Written Examination Callaway Plant Reactor Operator pressurizer pressure. C. Incorrect. 455 fails open. 2/4 high pressure B/S logic satisfied resulting in a reactor trip on high pressurizer pressure. D. Incorrect. PORV does open because 455 high trips one of the B/Ss and the controlling channel 456 trips the remaining B/S.

Technical Reference(s

): Dwg 7250D64-S006, Pressurizer Trip Signals, and S017, Pressurizer Pressure Relief System (Train A)

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, Analyz e the correct pr ocedure(s) to be utilized and the required actions to stabilize the plant, and Obj E, Identify the conditions that would require a Reactor Trip/Turbine Trip in OTO-BB-00006, Pressurizer Pressure Control Malfunction.

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: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 012 - Reactor Protection System Group # 1 K/A # 012 K2.01 Importance Rating 3.3 Knowledge of bus power supplies to the following: RPS channels, components, and interconnections.

Question #37

Which of the following supplies power to SSPS 'B' input relays?

Power is supplied by-A. NN02 only.

B. NN04 only.

C. NN02 and NN04 only.

D. NN01, NN02, NN03 and NN04.

Answer: D Explanation:

A. Incorrect. See explanation for 'D'. B. Incorrect. See explanation for 'D'. C. Incorrect. See explanation for 'D'.

D. Correct. There are 4 channels for each SSPS Trai n, therefore each SSPS Tr ain has a power input from each SR 120v AC NN bus.

Technical Reference(s

): E-23NN01, Class 1E Instrument AC Schematic

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-27, Obj G, State the power source to the RPS and to the Reactor Trip and Bypass breakers.

Question Source: Bank # __L15152____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 012 - Reactor Protection System Group # 1 K/A # 012 A2.06 Importance Rating 4.4 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Failure of RPS signal to trip the reactor.

Question #38

Given the following plant conditions:

• Reactor Power is 100%

• A Reactor Coolant Pump trips

• The Reactor fails to trip

• A Reactor trip is unsuccessful from SB HS-1, Reactor Trip Man Actuation, on RL003

• A transition is made to FR-S.1, Response To Nuclear Power Generation/ATWS In accordance with FR-S.1, what is the first action to be taken by the Reactor Operator?

A. Manually trip the turbine B. Insert control rods at maximum rate C. Manually trip the reactor using SB HS-1 D. Open supply break ers to PG19 and PG20 Answer: C Explanation:

A. Incorrect. Plausible as this is action taken in FR-S.1, Step 2 RNO. B. Incorrect. Plausible as this action is taken if tripping the reactor from SB HS-1 is unsuccessful. C. Correct. This is the first action to be taken by the RO in FR-S.1, Step 1 RNO. D. Incorrect. Plausible as this action is also taken in Step 1 RNO if manually tripping the reactor is not successful.

Technical Reference(s

): FR-S.1, Response To Nuclear Power Generation References to be provided to applicants during examination:

None NRC Site-Specific Written Examination Callaway Plant Reactor Operator Learning Objective: T61.003D, LP D-29, Obj C, State and Explain the Immediate Action steps, including Response Not Obtained (RNO) actions, of FR-S.1, Response to Nuclear Power Generation/ATWS.

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: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 013 Engineered Safety Features Actuation System Group # 1 K/A # 013 K1.12 Importance Rating 4.1 Knowledge of the physical connections and/or cause effect relationships between the ESFAS and the following systems: ED/G.

Question #39

Given the following plant conditions:

• A steamline break occurs in Containment

• Containment pressure stabilizes at 10 psig

Which one of the following will occur?

A. NE01, Emergency Diesel Generator 'A

', will start B. PEN01B, Containment Sp ray Pump 'B', will start C. PEG01C, Component Cooling Water Pump 'C', will start D. SGN01D, Containment Cooler

'D', will shift to fast speed

Answer: A Explanation:

A. Correct. NE01 receives a start signal on a Safety Injection Signal which occurs at 3.5 psig on containment pressure. B. Incorrect. Containment Spray pumps only start on a CSAS, not a Safety Injection Signal. C. Incorrect. CCW pump 'A' will start but CCW pump 'C' will not receive a start signal unless 'A' pump does not start. D. Incorrect. Containment coolers shift to slow speed on a SI signal, not fast speed.

Technical Reference(s

): E-23KJ01A, Diesel Generator KKJ01A Engine Control (Start/Stop Circuit)

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-3, Obj F, List and Describe the methods of generating a Standby NRC Site-Specific Written Examination Callaway Plant Reactor Operator Diesel Generator Start Signal.

Question Source: Bank # ______ Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 022 - Containment Cooling System Group # 1 K/A # 022 A4.05 Importance Rating 3.8 Ability to manually operate and/or monitor in the control room: Containment readings of temperature, pressure, and humidity system.

Question #40

Which one of the following Cont ainment Cooling System indications would require the operator to use adverse c ontainment values when responding to plant events in accordance with the Emergency Operating Procedures?

A. Containment Humidity, GN AI-27 92%

B. Containment Temperature, GN TR-63 125°F C. Containment Pressure, GN PI-938 3.8 psig D. Containment Radiation, GN RR-60 5X10E3 R/hour Answer: C Explanation:

A. Incorrect. High humidity would be an indication of a possible LOCA or secondary fault, but does not require use of adverse numbers in the EOPs.

B. Incorrect. This value does exceed the containment TS value that could require action, but is not used by the EOPs. C. Correct. Containment pressure over 3.5 requires adverse numbers to be applied during EOP operations. D. Incorrect. This is a parameter used for adverse conditions but is not required for use until radiation exceeds 10E5 R/hour.

Technical Reference(s

): E-0, Reactor Trip or Safety Injection References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-4, Obj I, State and Explain the criteria that determine Adverse Containment including the basis for each of the Criteria.

Question Source: Bank # ______ Modified Bank # ______

New ___X____

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 026 Containment Spray System Group # 1 K/A # 026 A3.01 Importance Rating 4.3 Ability to monitor automatic operation of the CSS, including: Pump starts and correct MOV positioning.

Question #41

Given the following plant conditions:

• Reactor power is 100%

• Containment Pressure Channel, GN PI-934, has failed

• ALL actions of OTO-GN-00001, Containment Pressure Channel Failure, have been completed

• Following completion of OTO-GN-00001, Co ntainment Pressure Channel, GN PI-936, fails HIGH Which of the following describes the stat us of the Containment Spray System?

A. Neither Containment Spray Pump is r unning; EN HV-1 and 7, Containment Recirc Sump To Containment Spray Pump Valves are OPEN

B. Both Containment Spray Pumps are runni ng; EN HV-1 and 7, Containment Recirc Sump To Containment Spray Pump Valves are OPEN

C. Both Containment Spray Pumps are running; BN HV-3 and 4, RWST Header To Containment Spray Pump Suction Valves are OPEN

D. Neither Containment Spray Pump is running; BN HV-3 and 4, RWST Header To Containment Spray Pump Suction Valves are OPEN Answer: D Explanation:

A. Incorrect. Correct for pump status. EN HV-1 and 7 will not be opened until the CS System is realigned for recirc following an actuation.

B. Incorrect. No pumps should start as there is only one input to the CSAS - that from GN PI-936. EN HV-1 and 7 will not be opened until the CS System is realigned for recirc following an actuation. C. Incorrect. No pumps should start as there is only one input to the CSAS - that from GN PI-936. BN HV-3 and 4 are open at the given power level. D. Correct. No pumps s hould start as there is only one input to the CSAS - that from GN PI-936. BN HV-3 and 4 are open at the given power level.

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

): M-22EN01, Containment Spray System OTO-SA-00001, Engineered Safety Feature Actuation Verification And Restoration References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 18, Obj G, Describe those conditions that will generate a containment spray actuation signal (CSAS) and Identify the Containment Spray System response to a CSAS.

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: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 026 Containment Spray System Group # 1 K/A # 026 2.4.49 Importance Rating 4.6 Ability to perform without reference to procedures those actions that require immediate operation of system component and controls.

Question #42

Given the following plant conditions:

• Reactor power is 100%

• Annun 59A, CSAS, alarms

• Containment pressure is normal Which one of the following actions is requi red to be performed IMMEDIATELY by the Reactor Operator?

A. Trip the Reactor B. Start both ESW Pumps C. Trip the Reactor Coolant Pumps D. Place the Containment Spray Pumps in Pull-To-Lock Answer: D Explanation:

A. Incorrect. Plausible as this would be prudent and required if cooling water cannot be restored to the RCPs in a timely manner. B. Incorrect. Plausible as this is action is incorporated in OTO-SA-00002, but it is not one of the Immediate Action steps. C. Incorrect. Plausible as the RCPs will lose cooling water. D. Correct. This is Step 1 of OTO-SA-00002 and is an Immediate Action step.

Technical Reference(s

): OTO-SA-00002, Spurious Containment Spray And Containment Phase B Isolation References to be provided to applicants during examination:

None NRC Site-Specific Written Examination Callaway Plant Reactor Operator Learning Objective: T61.003B, LP B-54, Obj F, Describe the immediate operator actions to respond to a spurious Containment Spray and Containment Phase B Isolation.

Question Source: Bank # ______ Modified Bank # ______ New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 039 - Main and Reheat Steam System Group # 1 K/A # 039 K4.04 Importance Rating 2.9 Knowledge of MRSS design feature(s) and/or interlock(s) which provide for the following: Utilization of steam pressure program control when steam dumping through atmospheric relief/dump valve, including T-ave. limits.

Question #43

Given the following plant conditions:

• Reactor power is 7%

• The Turbine is not synchronized to the grid

• RCS Tavg is being maintained by the Condenser Steam Dumps

• AB PK-507, Steam Hdr Press Ctrl, fails to zero

The Atmospheric Steam Dumps (ASDs) will start to open at

___(1)_____ and RCS Tavg will be ___(2)_____ when the ASDs are full open.

(1) (2) A. 1125 psig 567°F B. 1185 psig 560°F C. 1125 psig 560°F D. 1185 psig 567°F

Answer: A Explanation:

A. Correct. ASDs start to open at 1125 and are full open at 1185 which would correspond to an RCS temperature of 567°F. B. Incorrect. Setpoint is for when ASDs will be full open and Tavg is temperature when the ASDs start to open. C. Incorrect. Setpoint is correct but Tavg is temperature when the ASDs start to open. D. Incorrect. Setpoint is for when ASDs will be full open. Tavg is correct.

Technical Reference(s

): OOA-RL-00004, Main Control Board Controllers and Potentiometers

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination:

Steam Tables Learning Objective: T61.0110, LP-20, Obj B, Describe the purpose and operation of the following Main Steam system components: 1. Main Steam Line (MSL) Safety and Atmospheric Steam Dumps (ASDs), including setpoints.

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: 55.41.4 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 059 - Main Feedwater System Group # 1 K/A # 059 K3.04 Importance Rating 3.6 Knowledge of the effect that a loss or malfunction of the MFW will have on the following: RCS.

Question #44

The following plant conditions exist:

• Reactor power is 100% at EOL

• A feedwater system transient results in high water levels in High Pressure Feedwater Heaters 7A and 7B

• Annun 102D, HP FW HTR LEV HIHI, alarms Which one of the following identifies the INITIAL effect on-

(1) Indicated RCS Tcold

AND

(2) Reactor power

A. (1) RCS Tcold lowers (2) Reactor power lowers

B. (1) RCS Tcold lowers (2) Reactor power rises

C. (1) RCS Tcold rises (2) Reactor power lowers

D. (1) RCS Tcold rises (2) Reactor power rises

Answer: B Explanation:

A. Incorrect. Reactor power rises.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Correct. At EOL the moderator temperature coefficient is negative. With hihi heater levels, extraction steam is lost resulting in colder feedwater which will result in RCS Tcold be colder. Colder Tcold with a negative moderator temperature coefficient results in positive reactivity added to the core, resulting in reactor power rising.

C. Incorrect. Tcold lowers and reactor power rises. D. Incorrect. Tcold will lower.

Technical Reference(s

): OTA-RK-00024 Addendum 102D, High Pressure Feedwater Heater Level High High References to be provided to applicants during examination:

None Learning Objective: T61.GFES, LP 40, Obj 4, Describe the effect on the moderator temperature coefficient of reactivity from changes in: c. core age. T61.0110, LP 32, Obj A, State the function of the following Subsystems 4. High Pressure (HP) Feedwater Heaters Subsystem.

Question Source: Bank # __X____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____Watts Bar 2009________

Question Cognitive Level

Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content

55.41.1 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 061 - Auxiliary / Emergency Feedwater System Group # 1 K/A # 061 K3.02 Importance Rating 4.2 Knowledge of the effect that a loss or malfunction of the AFW will have on the following: S/G.

Question #45

Given the following plant conditions:

• A plant trip occurs from 100% power

• A lockout of ESF Train 'B' 4160 Volt Bus NB02 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. 'B' and 'D' Answer: C Explanation:

A. Incorrect. AFW Pump 'B' supplies 'A' and 'D'. NB02 supplies power to AFW Pump 'B'. B. Incorrect. Supplied by MDAFW Pump 'B'.

C. Correct. These 2 SGs are supplied by MDAFW Pump 'A', which still has power. D. Incorrect. 'B' MDAFW Pump supplies 'D'; 'B' would be supplied by MDAFW Pump 'A'.

Technical Reference(s

): M-22AL01, Auxiliary Feedwater System References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-25, Obj C, Draw, Label and Explain a one line diagram of the AFW System to include the components listed in Objective B showing system alignment under the following conditions: 1. Normal operations 3. Auxiliary Feedwater Actuation Signal (AFAS).

Question Source: Bank # __X____

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

Question History: Last NRC Exam ___Callaway 2007 Exam_________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.4 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 061 - Auxiliary / Emergency Feedwater System Group # 1 K/A # 061 K5.02 Importance Rating 3.2 Knowledge of the operational implications of the following concepts as they apply to the AFW: Decay heat sources and magnitude.

Question #46

Given the following plant conditions:

• The plant is in Mode 3 following a reactor trip from 100%

• IAW OTG-ZZ-00006, Plant Cooldown Hot Standby To Cold Shutdown, a plant cooldown of 50°F/hr is commenced using Auxiliary Feedwater (AFW)

• Steam Generator (SG) Narrow Range levels are maintained stable at 50% during the cooldown

Which one of the following descri bes the trend of Auxiliary Feedwater flow requirements as the plant cools down to Mode 5?

A. AFW flow requirements are constant as long as SG level remains constant B. AFW flow requirements are constant as long as the cooldown rate remains constant C. More AFW flow is required to maintain SG level due to a rise in SG water density as it cools D. Less AFW flow is required to maintain SG level because heat input to the SGs lowers as the cooldown continues Answer: D Explanation:

A. Incorrect. Plausible if heat input to the SG did not change. Heat input lowers due to less decay heat as the cooldown progresses.

B. Incorrect. Plausible if the effects of less decay heat is not considered. C. Incorrect. Water density does not have any impact at this temperature in the SG. D. Correct. Heat input lowers as the cooldown progresses due to less decay heat from the reactor.

Technical Reference(s

): OTG-ZZ-00006, Plant Cooldown Hot Standby To Cold Shutdown

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination:

None Learning Objective: T61.GFES, LP-44 (Chap 8), Obj 24, Define decay heat, and Obj 25, Explain the relationship between decay heat generation and: a. Power level history b. Power production c. Time since reactor shutdown.

T61.0110, LP-25, Obj A, State the function of the Auxiliary Feedwater (AFW) System.

Question Source: Bank # __R12588____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.5 Comments: Question examines knowledge of AFW flow and decay heat. Past plant events have resulted from operators failing to adequately control AFW flow post trip, causing excessive cooldown. This is a common misconception as indicated by the exam history of this question in LOCT having a success rate of 71%.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 062 - A.C. Electrical Distribution Group # 1 K/A # 062 K4.01 Importance Rating 2.6 Knowledge of ac distribution system design feature(s) and/or interlocks(s) which provide for the following: Bus lockouts.

Question #47

Which one of the following would result in a bus lockout to NB02?

A. An Overvoltage condition B. An Undervoltage condition C. A Timed Ground Overcurrent condition

D. A XNB02 Transforme r Lockout condition Answer: C Explanation:

A. Incorrect. This condition does not result in any lockout or loss of power. B. Incorrect. Results in a loss of power to the bus. There is no bus lockout associated with this condition and the EDG would start and repower the bus.

C. Correct. This condition will result in a loss of power to the bus with a bus lockout preventing the ability to reenergize the bus. D. Incorrect. This condition will result in a transformer lockout and a loss of power to the bus. A bus lockout does not occur and the bus can be reenergized from the EDG.

Technical Reference(s

): OTA-RK-00016, Addendum 21A, NB02 Bus Lockout

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-6, Obj A, Draw and Explain a one line diagram of the Safeguards Power system to include the components and subsystems listed in objective B with normal breaker alignments shown.

Question Source: Bank # ______ Modified Bank # ______ New ___X____

Question History: Last NRC Exam ____________

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 062 - A.C. Electrical Distribution Group # 1 K/A # 062 A1.01 Importance Rating 3.4 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the ac distribution system control including: Significance of D/G load limits.

Question #48

A diesel generator (DG) monthly surveillance is being performed in accordance with OSP-NE-0001A, Standby Diesel Generator A Periodic Tests.

Which one of the following par ameters would render the te st results unacceptable?

A. DG obtained speed and voltage in 11.8 seconds

B. DG MW load was maintained at 6.3 MW during the surveillance C. Fuel oil storage tank leve l was recorded as 80,980 gallons D. B' starting air tank pressure was recorded as 635 psig Answer: B Explanation:

A. Incorrect. Required speed and voltage must be obtain within 12 seconds. B. Correct. The DG is required to be loaded with 5.6 to 6.2 MW for 60 minutes. C. Incorrect. Minimum fuel oil storage tank level is 80,900 gallons (79.43%). D. Incorrect. At least one starting air tank must be greater than 610 psig.

Technical Reference(s

): OSP-NE-0001A, Standby Diesel Generator A Periodic Tests References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-3, Obj M, State the applicable Technical Specifications for the Standby Diesel Generators.

Question Source: Bank # ______ Modified Bank # ______ New ___X____

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.8 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 063 - D.C. Electrical Distribution Group # 1 K/A # 063 A4.03 Importance Rating 3.0 Ability to manually operate and/or monitor in the control room: Battery discharge rate.

Question #49

An earthquake has caused a Loss of All AC Power. NK12 battery discharge amps are at 200 amps.

Which one of the following is the MAXIMUM time that NK12 could supply NK02 bus? (Assume NK12 was fully charged at the time of the earthquake.)

A. 12.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> B. 6.25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br /> C. 6.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> D. 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />

Answer: D Explanation:

A. Incorrect. Value could be obtained by using battery capacity for PK batteries (2400 AH). B. Incorrect. Value could be obtained by using battery capacity for NK01 and NK04 (1250 AH). C. Incorrect. Value could be obtained by using PJ battery capacity (1200AH). D. Correct. Rating is equal to Rating/Discharge Rate (900 AH/200 A = 4.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

Technical Reference(s

): E-21NK02, Class 1E 125V DC System Meter & Relay Diagram References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-6, Obj B, Describe the purpose and operation of the following Safeguards Power System components and subsystems: 5. 125 VDC System (NK).

Question Source: Bank # __0110060B02A (L4521)____ Modified Bank # ______

New _______

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question History: Last NRC Exam ___Callaway 1998 Retake_________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.8 Comments:

Two distractors changed to make credible calculations.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 064 - Emergency Diesel Generators Group # 1 K/A # 064 A2.02 Importance Rating 2.7 Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Load, VARS, pressure on air compressor, speed droop, frequency, voltage, fuel oil level, temperatures.

Question #50

Given the following plant conditions:

• Reactor power is 100%

• NE02, 'B' Emergency Diesel Generator (EDG), has been started IAW OSP-NE-0001B, Standby Diesel Generator B Periodic Tests, and is connected to ESF Train 'B' 4160 Volt Bus NB02

• NB0209, NB02 Norm Sply Bkr, is closed

• ANN 23C, DG NE02 OV, alarms Which one of the following actions should be taken by the crew in response to the above conditions?

A. Ensure Parallel Operation Light is LIT on Local Panel NE106 and lower voltage using NE HS-14A, DG NE02 Auto Volt Reg

B. Place NE HS-5, DG NE01 Unit Parallel, in Reset to acti vate the speed droop circuit for NE02 and lower voltage using NE HS-14A, DG NE02 Auto Volt Reg C. Trip NE02, 'B' Emergency Diesel G enerator, and enter OT O-NB-00002, Loss of Power to NB02

D. Open NB0209, NB02 Norm Sply Bkr, to a llow the Static Exciter Voltage Regulator (SEVR) to control the EDG in the isochronous mode

Answer: A Explanation:

A. Correct. Action is correct IAW Annun 23C response to clear the OV condition. B. Incorrect. Placing NE HS-5 to reset will remove the speed droop circuit from service, not activate it.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Incorrect. If voltage cannot be reduced, the OTA directs to operator to secure the EDG rather than trip it. Also, the entry conditions to OTO-NB-00002 would not be met with NB0209 closed. D. Incorrect. The operating mode for the SEVR is determined by the Unit Parallel Switch, not the position of the bus feeder breakers.

Technical Reference(s

): OTA-RK-00016 Addendum 23C, Diesel Generator NE02 Overvoltage References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-3, Obj F, List and Describe the methods of generating a Standby Diesel Generator Start Signal.

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

55.41.8 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 064 - Emergency Diesel Generators Group # 1 K/A # 064 2.4.45 Importance Rating 4.1 Ability to prioritize and interpret the significance of each annunciator or alarm.

Question #51

Given the following plant conditions:

• NE02, 'B' Emergency Diesel Generator, has started due to a loss of power to NB02

• Annun 23D, DG NE02 TROUBLE, alarms

• The Secondary OT reports the following alarms are lit on the local annunciator panel, Panel KJ-122, for NE02:

• 2B, LUBE OIL LEVEL LOW

• 2E, LUBE OIL TEMP HIGH

• 4A, JACKET WATER PRESS LOW

• 6F, CRANKCASE PRESS HIGH

Which alarm will have the highest priority?

A. 2B, LUBE OIL LEVEL LOW B. 2E, LUBE OIL TEMP HIGH C. 4A, JACKET WATER PRESS LOW D. 6F, CRANKCASE PRESS HIGH

Answer: D Explanation:

A. Incorrect. This condition will not trip the EDG. B. Incorrect. This condition will not trip the EDG. C. Incorrect. This condition will not trip the EDG. D. Correct. If not addressed, high crankcase pressure will trip the EDG.

Technical Reference(s

): OTA-KJ-00122 Addendum 6F, Crankcase Pressure High

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP-3, Obj J, List all conditions which will cause a Diesel Engine to Trip (Stop).

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: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 073 - Process Radiation Monitoring System Group # 1 K/A # 073 2.1.18 Importance Rating 3.6 Ability to make accurate, clear, and concise logs, records, status boards, and reports.

Question #52

Given the following plant conditions:

• Reactor power is 100%

• A loss of NB01 occurs

What impact does the above have on the Pr ocess Radiation Monitoring System?

A. MCB Annunciators 61A, PROCESS RAD HIHI, and 61B, PROCESS RAD HI, will not annunciate.

B. Communication has been lost between A LL radiation monitors and RM-11, Radiation Monitor Control Panel.

C. No plant discharge can be initiated due to the loss of HB RE-18, Radwaste Bldg Discharge Line Rad Monitor, IAW HTP-ZZ-02006, Liquid Radwaste Release Permit (Batch).

D. GK RE-04, Control Room Ventilation Rad Monitor, log readings must be initiated IAW OSP-ZZ-00001, Control Room Shi ft and Daily Log Readings and Channel Checks.

Answer: D Explanation:

A. Incorrect. MCB annunciators will still function. B. Incorrect. All rad monitors not affected. Some have had a loss of communication and others have not been affected.

C. Incorrect. A plant discharge can always be initiated. With a loss of HB RE-18 a discharge can be initiated if two independent samples are analyzed and verified (FSAR Table 16.11-2, Action 31). D. Correct. Only a loss of communication has occurred with the RM11 for select rad monitors. The RM23 can still be used to monitor the affected rad monitors and maintain operability.

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

): References to be provided to applicants during examination:

None Learning Objective:

T61.0110, LP-36, Obj D, List the systems that interface with the Process Radiation Monitoring System and Explain how a loss of power affects the Process Radiation Monitoring Systems.

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: 55.41.11 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 076 - Service Water System Group # 1 K/A # 076 K1.01 Importance Rating 3.4 Knowledge of the physical connections and/or cause-effect relationships between the SWS and the following systems: CCW system.

Question #53

Given the following plant conditions:

• The plant is in Mode 5

• Both Component Cooling Water (CCW) Trains are in service to support Residual Heat Removal (RHR) System dual train operation

• Both Essential Water System (ESW)

Trains are in their standby lineup

• Emergency Diesel Generator NE01 is Out of Service for preplanned maintenance

• NB0112, NB01 Norm Sply Bkr, trips OPEN Which one of the following describes the effect the above conditions will have on the CCW System?

Cooling water will be lost to: A. BOTH CCW Trains B. ONLY CCW Train 'A' C. ONLY CCW Train 'B' D. NEITHER CCW Train Answer: D Explanation:

A. Incorrect. Neither CCW train will lose flow - see explanation for D. B. Incorrect. Neither CCW train will lose flow - see explanation for D. C. Incorrect. Neither CCW train will lose flow - see explanation for D. D. Correct. Flow is maintained to 'A' train by service water since the cross-connect valves will have no power and will fail as is (open) on the loss of NB01. Service water will continue to provide cooling water NRC Site-Specific Written Examination Callaway Plant Reactor Operator flow to CCW Train 'B'.

Technical Reference(s

): M-22EF01, Essential Service Water System M-22EF02, Essential Service Water System References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 5, Obj C, Describe the operation of the Essential Service Water System under the following conditions: 3. Loss of Offsite Power; 5. Opposite train NB bus undervoltage with low flow to the CTMT coolers.

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: 55.41.8 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 078 - Instrument Air System Group # 1 K/A # 078 A3.01 Importance Rating 3.1 Ability to monitor automatic operation of the IAS, including: Air pressure.

Question #54

Given the following plant conditions:

• Reactor power is 100%

• KA HS-43, COMP SEQ SEL SW, is in t he C-A-B position for the Compressed Air System

• An instrument air line break occurs at the condensate polishers

• Compressed air system pressure lowers to 114 psig and stabilizes Which one of the following de scribes the status of t he Compressed Air System? A. Only Air Compressors 'A' and 'C' are loaded; KA PV-11, Service Air Sply Vlv, is CLOSED B. Only Air Compressors 'A' and 'C' are loaded; KA PV-11, Service Air Sply Vlv, is OPEN C. All Air Compressors are loaded; KA PV-11, Service Air Sply Vlv, is OPEN D. All Air Compressors are loaded; KA PV-11, Service Air Sply Vlv, is CLOSED Answer: C Explanation:

A. Incorrect. All compressors should be loaded when pressure lowers to 115 psig; KA PV-11 does not close until 110 psig.

B. Incorrect. All compressors should be loaded when pressure lowers to 115 psig; KA PV-11 does not close until 110 psig. C. Correct. All compressors should be loaded when pressure lowers to 115 psig; KA PV-11 does not close until 110 psig. D. Incorrect. KA PV-11 does not close until 110 psig.

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

): OTO-KA-00001, Partial Or Total Loss Of Instrument Air OTA-RK-00024 Addendum 92A, Compressed Air Pressure Low

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 and Obj F, Identify the Service and Instrument Air System Main Control Board (MCB) controls, alarms and indications and Describe how each is used to predict, monitor or control changes in the Service and Instrument Air System.

Question Source: Bank # __R12189____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.4 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 103 - Containment System Group # 1 K/A # 103 A2.05 Importance Rating 2.9 Ability to (a) predict the impacts of the following malf unctions or operations on the containment system and (b) based on those predictions, use procedure to correct, control, or mitigate the consequences of those malfunctions or operations: Emergency containment entry.

Question #55

What is the requirement for the Emergency Personnel Hatch (EPH) in order to make an Emergency Containment Entry in accordance with HDP-ZZ-06100, Reactor Building Access?

A. The EPH Outside Door Shield MUST be removed prior to personnel entering Containment

B. The EPH Outside Door Shield MUST be removed within one hour of personnel entering Containment

C. Removal of the EPH Outside Door Shield can be WAIVED by the Emergency Duty Officer D. Removal of the EPH Outside Door Shield is NOT a prerequisite for entering the Containment

Answer: D Explanation:

A. Incorrect. Plausible as this would be the most conservative action for a RB entry. B. Incorrect. It must be removed expeditiously as practical, but no specific time is given for its removal. C. Incorrect. No provision allowed for this action to be waived. D. Correct. Per a note in HDP-ZZ-06100, the EPH outside door shield is not a prerequisite to RB entry, but it should be removed as expeditiously as practical.

Technical Reference(s

): HDP-ZZ-06100, Reactor Building Access

References to be provided to applicants during examination:

None Learning Objective: T61.003A, LP A-8, Obj B, Discuss the following as they apply to HDP-ZZ-06100,Reactor Building Access: 5. Discuss the process for an emergency containment entry.

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

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content

55.41.12 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 002 - Reactor Coolant System Group # 2 K/A # 002 2.4.11 Importance Rating 4.0 Knowledge of abnormal condition procedures.

Question #56

Given the following plant conditions:

• Reactor power is 100%

• Charging Flow - 150 gpm

• Letdown Flow - 75 gpm

• 'A' Centrifugal Charging Pump (CCP) in service

• Annun 32C, PZR LO LEV DEV, is in alarm

• Pressurizer (PZR) Level - STABLE In accordance with OTO-BB-00003, RCS Excessive Leakage, which one of the following describes actions that will be implem ented by the Control Room Crew for plant conditions? A. Start 'B' CCP, monitor PZR level, conti nue Power Operation if PZR level rises to program level

B. Trip the reactor; go to E-0, Reactor Trip or Safety Injection, Safety Inject if PZR pressure approaches the SI setpoint

C. Enter OTO-MA-00008, Rapid Load Reduction , to place the plant in Mode 3 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> due to excessive RCS leakage

D. Reduce letdown flow to 45 gpm, perfo rm OSP-BB-00009, RCS In ventory Balance, and trip the reactor if PZR level continues to lower Answer: B Explanation:

A. Incorrect. Plausible as this action could stabilize PZR level, but continued plant operation is not allowed with this amount of RCS leakage.

B. Correct. Correct action as OTO-BB-00003 requires a reactor trip if RCS leakage is calculated to exceed 50 gpm.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator C. Incorrect. Plausible as the 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> limit is the correct time to reduce RCS leakage if the TS limit is exceeded. The leakage amount in the given conditions require a plant trip and not a controlled shutdown. D. Incorrect. Plausible as reduced letdown could allow PZR level to stabilize, but a reactor trip is required per the OTO if the leak rate exceeds 50 gpm.

Technical Reference(s

): OTO-BB-00003, RCS Excessive Leakage References to be provided to applicants during examination:

None Learning Objective: T61.003B, LP B-12, Obj F, Using the plant computer or data provided by the instructor, Determine the amount of excessive RCS leakage. Confirm the leakage rate using charging and letdown mismatch and Classify the leak rate in terms of the required Plant Shutdown and Cooldown.

Question Source: Bank # __R12178____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 011 - Pressurizer Level Control System Group # 2 K/A # 011 K2.01 Importance Rating 3.1 Knowledge of bus power supplies to the following: Charging pumps.

Question #57

Given the following plant conditions:

• Reactor power is 100%

• A loss of offsite power occurs

• NE02, 'B' Emergency Diesel Generator, fails to start Which one of the following co mponents should be in service? (Assume NO operator action has taken place)

A. 'A' Air Compressor B. 'A' Centrifugal Charging Pump C. 'C' Component Cooling Water Pump D. 'B' Pressurizer Backup Heater Group

Answer: B Explanation:

A. Incorrect. The SD load shed signal has shed power from 'A' air compressor. Power could be restored but no operator action has been taken to do so. B. Correct. Operator needs to know that NB01 has power and that 'A' CCP is powered from NB01. C. Incorrect. 'C' CCW pump has power but would not receive a start signal with the conditions given.

D. Incorrect. 'B' Heater Group is powered from NB02 and this bus has no power with the conditions given.

Technical Reference(s

): E-21005, List Of Loads Supplied By Emergency Diesel Generator References to be provided to applicants during examination:

None Learning Objective:

T61.0110, LP 6, Obj I, List the loads supplied by NB01 and NB02.

Question Source: Bank # ______

NRC Site-Specific Written Examination Callaway Plant Reactor Operator 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: 55.41.8 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 014 - Rod Position Indication System Group # 2 K/A # 014 K5.01 Importance Rating 2.7 Knowledge of the operational implications of the following concepts as they apply to the RPIS: Reasons for differences between RPIS and step counter.

Question #58

Which one of the following describes the differences, and associated reasons, of the Bank Demand Position Indication System (S tep Counters) and the Digital Rod Position Indication (DRPI) System?

A. (1) The Step Counters are more accura te because they use actual demand pulses sent to the rods. (2) DRPI is more reliable because it uses actual demand pulses sent to the rods.

B. (1) The Step Counters are more accura te because they use actual demand pulses sent to the rods. (2) DRPI is more reliable because it uses proximity coils to determine control rod position.

C. (1) The Step Counters are more reliable because they use proximity coils to determine control rod position.

(2) DRPI is more accurate because it uses proximity coils to determine control rod position.

D. (1) The Step Counters are more reliable because they use proximity coils to determine control rod position.

(2) DRPI is more accurate because it uses actual demand pulses sent to the rods.

Answer: B Explanation:

A. Incorrect. Step counter accuracy reason is correct but DRPI reliability description is inaccurate. B. Correct. Both statements are true concerning a comparison of reliab ility and accuracy and the reasons for the rod control system step counters and DRPI. C. Incorrect. Step counters are less reliable than DRPI while DRPI is less accurate than the step counters. D. Incorrect. Proximity coils are used on DRPI with demand pulses used by the step counters.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator NOTE: Step Counters accuracy is

+/- 1 step (+/- 5/8 inch). DRPI accuracy is

+/- 4 steps (+/- 2.5 inches).

Technical Reference(s

):

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 26, Obj R, Explain how demand position indication is developed, and Obj S, Explain how DRPI full and half accuracy are developed.

Question Source: Bank # __X____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____Watts Bar 2003________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.6 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 016 - Non-Nuclear Instrumentation System Group # 2 K/A # 016 K3.12 Importance Rating 3.4 Knowledge of the effect that a loss or malfunction of the NNIS will have on the following: S/G.

Question #59

The plant was operating at 100% power when the controlling Steam Generator (SG)

Level Channel on 'A' SG FAILS to 100%.

If NO operator action is taken, which one of the following describes the expected plant response?

A. The reactor will TRIP on Lo-Lo SG Level.

B. The reactor will TRIP due to a Turbine Trip.

C. Feedwater flow to 'A' SG will initially LOWER, then RISE causing SG level to stabilize BELOW program level.

D. Feedwater flow to 'A' SG will initially RISE, then LOWER causing SG level to stabilize ABOVE program level.

Answer: A Explanation:

A. Correct. This malfunction will result in the FRV closing which will cause SG to lower and the reactor to trip at a SG level of17%.

B. Incorrect. The reactor will trip but not from a turbine trip. The reactor will trip on SG low level and the turbine will trip on the reactor trip. C. Incorrect. True about the S/G level lowering, but it will not stabilize. SG level will continue to lower until the reactor trips.

D. Incorrect. SG level will lower, not rise. The operator must understand Steam Generator Water Level Control to know what will happen with SG level on a level instrument failure.

Technical Reference(s

): OTO-SA-00001, Engineered Safety Feature Actuation Verification And Restoration References to be provided to applicants during examination:

None NRC Site-Specific Written Examination Callaway Plant Reactor Operator Learning Objective: T61.003B, LP B-40, Obj D, Given a set of plant conditions or parameters indicating a Steam Generator Water Level Control Malfunction, Analyze the correc t procedure(s) to be utilized and the required actions to stabilize the plant.

Question Source: Bank # _R8659_____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.4 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 System Group # 2 K/A # 033 K4.05 Importance Rating 3.1 Knowledge of design feature(s) and/or interlock(s) which provide for the following: Adequate SDM (boron concentration).

Question #60

Given the following plant conditions:

• The plant has just completed refueling the reactor

• A fuel storage pool verification has NOT been performed Spent Fuel Pool minimum requi red boron concentration is ___(1)

___ ppm.

Spent Fuel Pool leakage is compens ated for by making up from ________(2)

________. A. (1) 2000 (2) reactor makeup water

B. (1) 2165 (2) reactor makeup water

C. (1) 2000 (2) blended flow from CVCS

D. (1) 2165 (2) blended flow from CVCS

Answer: D Explanation:

A. Incorrect. Boron concentration is plausible as this is the minimum concentration for the Refueling Pool during refueling operations. B. Incorrect. Boron concentration is correct but makeup method is incorrect. Primary grade water would be used to compensate for evaporation, not leakage. C. Incorrect. Boron concentration is plausible as this is the minimum concentration for the Refueling Pool during refueling operations. Blended flow for leakage compensation is correct. D. Correct. Amount of boron given is required to ensure SDM and leakage is made up from blended flow to compensate for the boron that is loss due to the leakage.

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

): Tech Spec 3.7.16, Fuel Storage Pool Boron Concentration References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 24, Obj G, List the systems that interface with the FPCCS and Explain how a loss of the interfacing system or a loss of the FPCSS affects the other, and Obj H, State the LCOs associated with the Fuel Pool Cooling and Cleanup System (FPCCS) Technical Specifications (T/S) and Final Safety Analysis Report (FSAR): 2. T/S 3.7.16.

Question Source: Bank # ___X___ Modified Bank # ______

New _______

Question History: Last NRC Exam ___North Anna 2008_________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.7 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 034 - Fuel Handling Group # 2 K/A # 034 A4.02 Importance Rating 3.5 Ability to manually operate and/or monitor in the control room: Neutron levels.

Question #61

The following plant conditions exist:

• The plant is in Mode 6 with re fueling operations in progress

• NN14, 7.5 KVA Inv Fed From NK0411, is out of service for replacement

• XNN06, Instrument Bus XFMR, is out of service for cleaning

• Power is lost to NN01, 120 VAC Inst Pnl Class IE Grp 1, and cannot be restored

Which one of the following actions is required?

A. Suspend operations that would cause intr oduction into the RCS of water at a boron concentration less than 2000 ppm.

B. Continue refueling operations usi ng Gamma Metrics Channel SE NIR-61, Post Accident PR/SR Recorder.

C. Immediately initiate action to restore at least one source range neutron flux monitor to operable status.

D. Suspend operations that would cause intr oduction into the RCS of water at a boron concentration less than current RCS boron concentration.

Answer: A Explanation:

A. Correct. With only 1 SR neuron flux monitor available, as given in the plant conditions, TS require that operations must be suspended of any conditions that could add coolant to the RCS at a boron concentration less than 2000 ppm. B. Incorrect. Plausible as this would be true if a Gamma-Metrics channel was available as they can be used as a substitute; however, with the given conditions there is no Gamma-Metrics channel available. C. Incorrect. Plausible as this is a TS requirement if there are no operable SRs. Operator needs to analyze that with the given conditions that SR channel 32 is still available. D. Incorrect. Plausible as this would ensure current RCS conditions are maintained. Tech Specs however, allow coolant makeup as long as it is above the TS requirement of 2000 ppm.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Instrumentation power supplies are as follows: NN01 - Source Range Channel 31 and Gamma-metrics Channel 60A NN02 - Source Range Channel 32 NN04 - Gamma-metrics Channel 61

Technical Reference(s

): Tech Spec 3.9.3, Nuclear Instrumentation OTO-NN-00001, Loss of Safety Related Instrument Power E-23NN01, Class 1E Instrument AC Schematic

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 28, Obj G, State the purpose, power supplies and indicator location and Explain the operation of the Gamma-metrics (Post Accident) Nuclear Instruments, and Obj H, State the LCOs associated with the following Nuclear Instrumentation Technical Specifications: 3. 3.9.3.

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: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 035 - Steam Generator System Group # 2 K/A # 035 K6.01 Importance Rating 3.2 Knowledge of the effect of a loss or malfunction on the following will have on the S/Gs: MSIVs.

Question #62

Given the following plant conditions:

• Reactor power is 21%

• AB HV-11, Steam Generator (SG) D MSIV, fails closed due to a spurious signal

Which of the following descri bes the INITIAL response of:

(1) Delta T in SG 'D'?

AND (2) Steam Pressure in SG 'D'?

(1) (2) SG 'D' Delta T SG 'D' Pressure A. RISES RISES B. RISES LOWERS C. LOWERS RISES D. LOWERS LOWERS Answer: C Explanation:

A. Incorrect. Loop delta T goes to near zero. B. Incorrect. Loop delta T goes to near zero.

C. Correct. In the affected loop, RCS delta T lowers to zero and S/G pressure rises because heat removal has been reduced to minimal. D. Incorrect. Steam pressure will rise.

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

):

References to be provided to applicants during examination:

None Learning Objective: T61.003B, LP SB-15, Obj H, Respond to closure of the Main Steam Isolation Valves.

Question Source: Bank # __X____ Modified Bank # ______ New _______

Question History: Last NRC Exam ___Watts Bar 2009_________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.5 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 2 041 - Steam Dump System and Turbine Bypass Control Group # 2 K/A # 041 A3.05 Importance Rating 2.9 Ability to monitor automatic operation of the SDS, including: Main steam pressure.

Question #63

Given the following plant conditions:

• Reactor power is 10%

• AB PK-507, Steam Hdr Press Ctrl, is set at 7.28 turns and in AUTO The Balance of Plant Operator (BOP) places the Main Generator in service and starts to raise load toward 100 MW.

Which one of the following accurately de scribes the expected response of the Condenser Steam Dumps?

A. Close to maintain Main Steam Pressure at 1092 psig B. Open to maintain RCS temperature at 557°F C. Close to maintain RCS temperature at 557°F D. Open to maintain Main Steam Pressure at 1092 psig Answer: A Explanation:

A. Correct. As power is raised, steam pressure lowers. At 10% power the steam dumps are in the steam pressure mode and will be controlled by steam pressure. As steam pressure lowers, the steam dumps will close to maintain steam pressure at the no-load setpoint of 1092 (7.28). B. Incorrect. As generator load rises, main steam pressure will lower. Steam dumps will close in auto to maintain steam pressure at 1092. RCS temperature of 557 is the Tsat for 1092; however, RCS temperature at this power will be maintained by operator action to withdraw the control rods. In the steam pressure mode, RCS temperature is not an input into the control circuitry.

C. Incorrect. As generator load rises, main steam pressure will lower. Steam dumps will close in auto to maintain steam pressure at 1092. RCS temperature of 557 is the Tsat for 1092; however, RCS NRC Site-Specific Written Examination Callaway Plant Reactor Operator temperature at this power will be maintained by operator action to withdraw the control rods. In the steam pressure mode, RCS temperature is not an input into the control circuitry. D. Incorrect. Plausible as the steam dumps will operate in auto to maintain steam pressure at 1092 in their selected mode for this plant condition. Operator needs to understand how the secondary and primary plant respond at this power level and the control circuitry of the condenser steam dumps.

Technical Reference(s

): OTG-ZZ-00003, Plant Startup Hot Zero Power To 30% Power - IPTE Dwg 7250D64, Sheet 10, Functional Diagram Steam Dump Control

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 20, Obj D, Identify all Main Steam, Steam Dump and S/G controls, alarms and indications and Describe how each is used to predict, monitor or control the Main Steam, Steam Dump and S/G System, and Obj H, Describe the two Steam Dump operational modes and Explain the operating signals sent to the Steam Dumps in each mode.

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

55.41.4 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 045 - Main Turbine Generator System Group # 2 K/A # 045 K1.18 Importance Rating 3.6 Knowledge of the physical connections and/or cause-effect relationships between the MT/G system and the following systems: RPS.

Question #64

Given the following plant conditions:

• Reactor power is 60%

• 'C' Main Feed Reg Valve fails open result ing in 'C' Steam Generator level rising to 92%

Which one of the following accurately describes the response of the Main Turbine/Generator Trip System and the R eactor Protection System (RPS) to this failure?

A. P-14, HI SG Level, will ENERGIZE the ELECTRICAL TRIP SOLENOID to trip the turbine;

P-4, Reactor Trip, enables the RPS to ENERGIZE the

reactor trip breakers shunt coil to trip the reactor.

B. P-4, Reactor Trip, will DE-ENERGIZE the ELECTRICAL TRIP SOLENOID to trip the turbine;

P-9, Turb Trip-Reactor Trip, enabl es the RPS to DE-ENERGIZE the reactor trip breakers shunt coil to trip the reactor.

C. P-14, HI SG Level, will ENERGIZE the MECHANICAL TRIP SOLENOID to trip the turbine;

P-9, Turb Trip-Reactor Trip, enables the RPS to ENERGIZE the

reactor trip breakers shunt coil to trip the reactor.

D. P-4, Reactor Trip, will DE-ENERGIZE the MECHANICAL TRIP SOLENOID to trip the turbine;

P-4, Reactor Trip, enables the RPS to DE-ENERGIZE the

reactor trip breakers shunt coil to trip the reactor.

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

A. Incorrect. P-14 energizes the mech trip solenoid, not the elec trip solenoid. P-4 is a result of the reactor trip, not an enabler for the reactor trip. B. Incorrect. For the given conditions, P-4 will not be energized until after the turbine trips. P-9 will energize, not de-energize the shunt coils. C. Correct. P-14 will energize the mech trip solenoid at 91% to trip the turbine and P-9 will enable the RPS to trip the reactor by energizing the shunt coil on the reactor trip breakers. D. Incorrect. For the given conditions, P-4 will not be energized until after the turbine trips.

Note: To correctly answer this question, the operator must have a thorough understanding of how both the Turbine Generator and RPS system uses Permissives to accomplish turbine and reactor trips.

Technical Reference(s

): OTN-AC-00001, Main Turbine And Generator Systems E-23SB10A, Reactor Trip Switchgear Train A

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 27, Obj D, List all the RPS Permissive Signals, including setpoints, coincidence and function, and Obj G, State the power source to the RPS And to the Reactor Trip and Bypass breakers. T61.0110, LP 38, Obj D, List the Main Turbine electrical and mechanical trips, including setpoints and coincidence where applicable, and Obj E, Explain the function and operation of the Emergency Trip system (ETS).

Question Source: Bank # ______ Modified Bank # __R8643____ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.4 Comments: Question modified to include response of RPS. Original question only evaluated knowledge of turbine trip system. Stem also modified to support RPS response.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Modified ID: R8643 The plant is operating at 60% power. A high Steam Generator (SG) level leads to a Turbine Trip and Reactor Trip.

Which ONE of the following correctly describes the operation of the Main Turbine Trip System?

A. P-14 HI SG Level will ENERGIZE the MECHANICAL TRIP SOLENOID to trip the turbine.

B. P-4 Reactor Trip will ENERGIZE the ELECTRICAL TRIP SOLENOID to trip the turbine.

C. P-14 HI SG Level will ENERGIZE the ELECTRICAL TRIP SOLENOID to trip the turbine.

D. P-4 Reactor Trip will DE-ENERGIZE the MECHANICAL TRIP SOLENOID to trip the turbine

ANSWER: A

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 001 Control Rod Drive System Group # 2 K/A # 001 A2.14 Importance Rating 3.7 Ability to (a) predict the impacts of the following malfunction or operations on the CRDS and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Urgent failure alarm, including rod-out-of-sequence and motion-inhibit alarms.

Question #65

Given the following plant conditions:

• Turbine load is being raised following a re fueling outage

• Annun 79A, ROD CTRL URG FAIL, alarms

• Annun 65D, T REF/T AUCT HI, alarms

• The Balance of Plant Operator stops turbine loading

• It is determined that a Pulser Failure has occurred in the Rod Control System (1) What is the impact on the Rod Control System?

AND

(2) What action is required to be taken?

A. (1) ALL rod motion is inhibited.

(2) Turbine load must be raised to clear Annun 65D

B. (1) Rod motion is ONLY i nhibited in the affected bank (2) Turbine load must be raised to clear Annun 65D

C. (1) Rod motion is ONLY i nhibited in the affected bank (2) Turbine load must be lowered to clear Annun 65D D. (1) ALL rod motion is inhibited.

(2) Turbine load must be lowered to clear Annun 65D

Answer: D Explanation:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator A. Incorrect. Turbine load would have to be lowered to clear Annun 65D. B. Incorrect. Impact on the rod control system is incorrect and action to clear Annun 65D is also incorrect. C. Incorrect. Correct action but impact on rod control is incorrect. D. Correct. A pulser failure generates a logic cabinet urgent failure which inhibits all rod motion. To clear Annun 65D, turbine load needs to be lowered IAW OTO-SF-00001, Rod Control Malfunctions, which would be entered for the urgent failure.

Technical Reference(s

): OTA-RK-00022, Addendum 79A, Rod Control Urgent Failure OTO-SF-00001, Rod Control Malfunctions

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 26, Obj M, List the causes of a "Rod Control Urgent Failure" and Explain the effects on the system. T61.003B, LP B-45, Obj D, Given a set of plant conditions or parameters indicating a Rod Control Malfunction, 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 ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.41.4 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 GENERIC Group # 1 K/A # 2.1.14 Importance Rating 3.1 Knowledge of criteria or conditions that require plant-wide announcements, such as pump starts, reactor trips, mode changes, etc.

Question #66

A plant announcement is procedurally REQUIRED for each one of the following EXCEPT when:

A. Starting Reactor Coolant Pump 'A' B. A loss of instrument air has occurred C. Annunciator 62B, AREA RAD HI, alarms D. Starting CGG02A, Fuel/Aux Bldg Emergency Exhaust Fan Answer: C Explanation:

A. Incorrect. Required per OTN-BB-00003, Reactor Coolant Pumps. B. Incorrect. Required per OTO-KA-00001, Partial or Total Loss of Instrument Air. C. Correct. Area Rad Hi does not require a plant announcement; Annun 62A, Area Rad HiHi, would require an announcement if the affected area was occupied. D. Incorrect. Required per OTN-GG-00001, Fuel Building HVAC System.

Technical Reference(s

): OTA-RK-00020, Addendum 62A, Area Radiation High High, and B, Area Radiation High References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 66, Obj B.1, Explain the following as they pertain to Operations Department Communications: 1. Addendum 6 of ODP-ZZ-00001 (which incorporates Nuclear Division Policy, "Guidelines for Verbal Communications".

Question Source: Bank # ______ Modified Bank # ______ New ___X____

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 3 GENERIC Group # 1 K/A # 2.1.1 Importance Rating 3.8 Knowledge of conduct of operations requirements.

Question #67

The Operating Shift compliment may be (1) less than the minimum required for a period of time not to exceed (2) hours in order to ac commodate an unexpected absence of on-duty shift crew members, provi ded immediate action is taken to restore the shift compliment.

(1) (2)

A. 2 1 B. 2 2 C. 1 2 D. 1 1 Answer: C Explanation:

A. Incorrect. The shift compliment can only be 1 less than the minimum, not 2 less. B. Incorrect. The shift compliment can only be 1 less than the minimum, not 2 less. C. Correct. Shift compliment and allowance for absence are correct IAW ODP-ZZ-00001.

D. Incorrect. Shift compliment is correct, but 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> are allowed to fill an absence rather than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

Technical Reference(s

): ODP-ZZ-00001, Operations Department - Code Of Conduct References to be provided to applicants during examination:

None Learning Objective:

T61.0110, LP-66, Obj A, Explain the following as applied in ODP-ZZ-00001, Operations Dept - Code of Conduct: 11. Discuss: b. Unexpected absence requirements regarding shift complement.

Question Source: Bank # _003A01A11B01_____ Modified Bank # ______

NRC Site-Specific Written Examination Callaway Plant Reactor Operator New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 GENERIC Group # 1 K/A # 2.1.44 Importance Rating 3.9 Knowledge of RO duties in the control room during fuel handling, such as responding to alarms from the fuel handling area, communication with the fuel storage facility, systems operated from the control room in support of fueling operations, and supporting instrumentation.

Question #68

A fuel shuffle is being performed in the Spent Fuel Pool in preparation for the start of Refuel.

Which one of the following alarms would requ ire the Reactor Operator to shift the running Fuel/Aux Building Normal Exhaust Fan to SLOW?

A. RED alarm on GG RE-28, Fuel/Aux Bl dg Radiation Detector , GAS channel (283)

B. YELLOW alarm on GG RE-27, Fuel/Aux Bl dg Radiation Detector , PART channel (271)

C. YELLOW alarm on GG RE-28, Fuel/Aux Bl dg Radiation Detector , GAS channel (283)

D. RED alarm on GG RE-27, Fuel/Aux Bldg Radiation Detector, IODINE channel (272)

Answer: A Explanation:

A. Correct. Actuation signals are only actuated on RED alarms on the GAS channel. Fan is shifted per OTA to balance HVAC flow following a FBVIS and CRVIS actuation.

B. Incorrect. Plausible as GG RE-27 will cause an actuation signal, but it is the wrong monitor channel and alarm status. C. Incorrect. Plausible as this is the proper channel for actuation signal (GAS), but a YELLOW alarm will not cause an actuation. The alarm must be RED to cause an actuation. D. Incorrect Plausible as this is a RED alarm, but actuations do not come from the IODINE channel.

Technical Reference(s

): OTA-SP-RM011, Radiation Monitor Control Panel RM-11 References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 39, Obj C, Describe the flowpath through the Aux Bldg, Control Bldg and Fuel Bldg HVAC Systems during the following evolutions: 2. Fuel Handling Accident (FBVIS).

Question Source: Bank # ______

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

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 3 GENERIC Group # 2 K/A # 2.2.43 Importance Rating 3.0 Knowledge of the process used to track inoperable alarms.

Question #69

In accordance with ODP-ZZ-00017, when is an orange Work Request Submitted sticker placed on an annunciator?

A. On a new annunciator until it becomes operational.

B. When an annunciator is defeated due to causing nuisance alarms.

C. When an annunciator is alarming due to an I&C surveillance procedure.

D. When a Job Request is submitted on a component affecting an annunciator.

Answer: D Explanation:

A. Correct. Would be part of a modification package. There would not be a requirement for any sticker on the annunciator. B. Incorrect. Green sticker would be used IAW Step 4.5.11 of ODP-ZZ-00017. C. Incorrect. Annunciator would not be defeated IAW ODP-ZZ-00017. D. Correct. Orange sticker would be used IAW Step 4.2 of ODP-ZZ-00017.

Technical Reference(s

): ODP-ZZ-00017, Annunciator Status And Tracking

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 74, Obj E, Discuss the following as they pertain to ODP-ZZ-00017, "Annunciator Status and Tracking": 3. Use of condition tags on annunciators associated with a Work Request. Question Source: Bank # __003A06D301A____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 GENERIC Group # 2 K/A # 2.2.13 Importance Rating 4.1 Knowledge of tagging and clearance procedures.

Question #70

In accordance with APA-ZZ-00310, Workman' s Protection Assurance, short duration maintenance, troubleshooting and testing can be performed under which one of the following?

A. caution tag B. hold off tag C. local control tag D. test in progress tag Answer: C Explanation:

A. Incorrect. Tag is used for information purposes only; operation is allowed if the information associated with the caution is adhered to and recognized. B. Incorrect. Tag used to protect personnel from work being performed on equipment. Equipment with a hold off tag can never be operated.

C. Correct. Rules of usage are correct for local control tags as given in Section 4.8 of APA-ZZ-00310. D. Incorrect. Tag is placed on handswitches in the CR for information purposes for CR operators.

Technical Reference(s

): APA-ZZ-00310, Workman's Protection Assurance References to be provided to applicants during examination:

None Learning Objective: T61.003A, LP A-13, Obj A.2.c, Describe the types of tags and their rules of usage. Local Control Tag Question Source: Bank # __003A33A202A____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 GENERIC Group # 2 K/A # 2.2.35 Importance Rating 3.6 Ability to determine Technical Specification Mode of Operation.

Question #71

The following plant conditions exist:

• Reactivity Condition (k eff) 0.98

• Average Reactor Coolant Temperature (°F) 325

• Reactor Vessel Head Closure Bolts Fully Tensioned 53 of 54 Which one of the followi ng Modes, as defined in Technical Specifications, is the plant in?

A. Mode 1 B. Mode 3 C. Mode 4 D. Mode 5

Answer: C Explanation:

A. Incorrect. Temperature would be greater than 350 and Keff would be greater than .99. B. Incorrect. Temperature would be greater than 350; Keff is correct. C. Correct. Given conditions describe Mode 4 IAW Tech Spec Table 1.1-1. D. Incorrect. Mode 5 temperature would be less than 200; Keff is correct.

Technical Reference(s

): Technical Specifications, Section 1, Use and Application; Definitions References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 77, Obj D, List the six operational Modes as defined in Technical Specifications.

Question Source: Bank # ______

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

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 3 GENERIC Group # 3 K/A # 2.3.4 Importance Rating 3.2 Knowledge of radiation exposure limits under normal or emergency conditions.

Question #72

A 20 year old male Health Physics contract or, with a complete exposure history, has already accumulated 500 MREM whole body To tal Effective Dose Equivalent (TEDE) dose this year at other nuclear plants.

Which one of the following is the maximum whole body TEDE dose that this worker can receive at the Callaway Plant this year without exceeding the administrative exposure limits of APA-ZZ-01000, Callaway Energy C enter Radiation Pr otection Program?

A. 500 MREM B. 1500 MREM C. 2000 MREM D. 3500 MREM Answer: C Explanation:

A. Incorrect. Plausible as this is a federal limit for a pregnant female. B. Incorrect. Plausible as this would be correct if you subtracted dose prior to coming to Callaway. The admin limit does not consider prior dose from another site unless it would result in exceeding a dose of 4 Rem. C. Correct. Admin limit is 2 Rem not to exceed 4 Rem including prior site. D. Incorrect. Plausible as this is an allowed limit (4000 - 500 = 3500) including prior sites, but only 2000 of that allowed limit can come from exposure at Callaway.

Technical Reference(s

): APA-ZZ-01000, Callaway Energy Center Radiation Protection Program

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 75, Obj C.8, APA-ZZ-010000, Callaway Plant Radiation Protection Program: 8. Discuss the Ameren Personnel exposure limits for the following categories: a. Total Effective NRC Site-Specific Written Examination Callaway Plant Reactor Operator Dose Equivalent (TEDE) b. Eye (LDE) c. Total Organ Dose Equivalent (TODE) d. Skin and Extremities (SDE).

Question Source: Bank # __L14631____ Modified Bank # ______ New _______

Question History: Last NRC Exam ___Callaway 2004 Exam_________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.12 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 GENERIC Group # 3 K/A # 2.3.13 Importance Rating 3.4 Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc.

Question #73

The following radiation levels are present in the Normal Charging Pump (NCP) Room due to a leak on the pump seal:

• Radiation 110 mrem/hr

• Contamination 1500 dpm/100 cm 2 beta-gamma

Which one of the radiological postings would be correct for the NCP Room?

A. Caution High Radiation Area Caution High Contaminated Area

B. Caution Radiation Area Caution Contaminated Area

C. Caution Radiation Area Caution High Contaminated Area

D. Caution High Radiation Area Caution Contaminated Area

Answer: D Explanation:

A. Incorrect. HCA would be >100,000 dpm/100 cm 2 beta-gamma. B. Incorrect. RA would be < 100 mr/hr. C. Incorrect. RA would be < 100 mr/hr and HCA would be >100,000 dpm/100 cm 2 beta-gamma. D. Correct. HRA is >100 mr/hr and <1000 mr/hr CA is >1000 dpm/100 cm 2 beta-gamma and <100,000

Technical Reference(s

): HDP-ZZ-01500, Radiological Postings 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 E.1.b/d, HDP-ZZ-01500, Radiological Posting, Define and Discuss the posting requirements of the following areas: High Radiation Area-HRA and Contaminated Area Question Source: Bank # ______ Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.12 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 1 Tier # 3 GENERIC Group # 4 K/A # 2.4.5 Importance Rating 3.7 Knowledge of the organization of the operating procedures network for normal, abnormal, and emergency evolutions.

Question #74

Which of the following statem ents describe proper use of t he Critical Safety Function Status Trees?

A. During functional restoration of a Core Cooling ORANGE path, a Heat Sink RED path develops. The Operator should continue with the func tional restoration of the Core Cooling ORANGE path.

B. During functional restoration of an Integrity ORANGE pa th, a Containment ORANGE path develops. The Operator should go to the Containment ORANGE path for functional restoration.

C. During functional restoration of a Heat Sink RED path, a Core Cooling ORANGE path develops. The Operator should continue with the func tional restoration of the Heat Sink RED path.

D. During functional restoration of a Heat Sink RED path, a Subcriticality ORANGE path develops. The Operator should go to the Subcriticality ORANGE path for functional restoration

Answer: C Explanation:

A. Incorrect. Though core cooling is a higher priority than heat sink, the red path associated with the heat sink would take priority and should be addressed immediately.

B. Incorrect. Integrity has a higher priority than containment and should be completed rather than implementing the containment path. C. Correct. A red path is a higher priority than any orange path. D. Incorrect. Plausible because the FRG procedures are correct; however, the EOPs and ECAs are part of the Optimal Recovery Guidelines.

Technical Reference(s

): ODP-ZZ-00025, EOP/OTO User's Guide

NRC Site-Specific Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-1, Obj K, Explain how challenges to critical safety functions are prioritized within each critical safety function.

Question Source: Bank # __L2256____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 GENERIC Group # 4 K/A # 2.4.14 Importance Rating 3.8 Knowledge of general guidelines for EOP usage.

Question #75

The following Step 25 of an EOP is in progress:

ACTION/EXPECTED RESPONSE RESPONSE NOT OBTAINED (RNO)

Establish Instrument Air To Containment

a. Ensure at least one service air compressor - RUNNING a. Start one service air compressor. b. Open Instrument Air Supply CTMT Iso Valve b. Dispatch an Operations Technician to verify air compressors and air dryers operating correctly. Refer to OTO-KA-00001, (PARTIAL LOSS OF INSTRUMENT AIR).

There are NO air compressors running. Step 25a RNO is being performed. NO air compressors can be started. By rules of usage, you should:

A. Go to 25b Action B. Go to 25b RNO C. Stay in 25a RNO D. Go back to 25a Action Answer: A Explanation:

A. Correct. Action is correct per Section 4.18 of ODP-ZZ-00025.

NRC Site-Specific Written Examination Callaway Plant Reactor Operator B. Incorrect. Step 25b rno would not be entered until 25b action step had been evaluated. C. Incorrect. Incorrect as you would not stay in a step that could not be performed unless the procedure specifically directed you not to proceed. D. Incorrect. Incorrect by rules of usage as this would result in a do loop in that you would keep going from 25a action to 25a rno.

Technical Reference(s

): ODP-ZZ-00025, EOP/OTO User's Guide

References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-1, Obj Q, Explain how the two column format is used for the following: 1. Action or expected condition is met 2. Action or expected condition is not met

3. Action or expected condition is not met and no contingency is provided 4. Action or expected condition is not met and contingency action is unsuccessful

Question Source: Bank # __L2246____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.41.10 Comments:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 1 008 - Pressurizer Vapor Space Accident Group # 1 K/A # 008 AA2.30 Importance Rating 4.3 Ability to determine and interpret the following as they apply to the Pressurizer Vapor Space Accident: Inadequate core cooling.

Question #76 Given the following plant conditions:

• Reactor Trip and Safety Injection have occurred

• RVLIS PUMPS OFF 10% and stable

• Core Exit Thermocouples 850 o F and rising

• PZR level 100%

• PZR pressure 1200 psig

• The operating crew has been unable to establish Safety Injection flow (1) What event has occurred?

AND

(2) What action will be directed by the Control Room Supervisor?

A. (1) A pressurizer vapor space break.

(2) Depressurize all Steam Generators IAW FR-C.1, Response to Inadequate Core Cooling B. (1) An RCS hot leg break.

(2) Depressurize all Steam Generators IAW FR-C.2, Response to Degraded Core Cooling C. (1) A pressurizer vapor space break. (2) Start Reactor Coolant Pumps IAW FR-C.1, Response to Inadequate Core

Cooling D. (1) An RCS hot leg break. (2) Start Reactor Coolant Pumps IAW FR-C.2, Response to Degraded Core Cooling Answer: A NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Explanation:

A. Correct. Action is correct IAW Step 11 of FR-C.1. RCPs would not be started unless RCS temperature rises above 1200°F. PZR level of 100% indicates that a pressurizer vapor space break has occurred. B. Incorrect. Indications are for entry into FR-C.1, not C.2. Directions by CRS are correct. C. Incorrect. Given conditions for pressurizer level indicate a vapor space leak rather than a RCS leak. Starting RCPs is an option in FR-C.1 but not until RCS temperature is above 1200°F. D. Incorrect. Given conditions for pressurizer level indicate a vapor space leak rather than a RCS leak. RCPs would not be started until RCS temperature is above 1200°F.

Technical Reference(s

): CSF-1, Critical Safety Function Status Trees (CSFST) FR-C.1, Response To Inadequate Core Cooling

References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-25, Obj A, Explain the Purpose and Major Action Categories of: 1. FR-C.1, Response To Inadequate Core Cooling and Obj B, Describe the Symptoms and/or Entry Conditions for: 1. FR-C.1, Response To Inadequate Core Cooling Question Source: Bank # __X____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____Callaway 2005 Exam________

Question Cognitive Level: Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

55.43.5 Comments: SRO level question based on the applicant having to assess the given plant conditions and selecting the correct strategy from the applicable FRG procedure to mitigate the inadequate core cooling condition.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 1 029 - Anticipated Transient Without Scram (ATWS) Group # 1 K/A # 029 2.4.18 Importance Rating 4.0 Knowledge of the specific bases for EOPs.

Question #77 Given the following plant conditions:

• Reactor Coolant Pump 'A' tripped from 100% power

• The reactor failed to trip and the crew has transitioned to FR-S.1, Response To Nuclear Power Generation/ATWS

• Reactor power is now 8% and stable

• At Step 14, Core Exit TCs are indicating 1250 o F and rising Which one of the following dire ctions will be given to the crew by the CRS and the reason for this direction?

A. Return to Step 4 of FR-S.1 because core damage is imminent.

B. Transition to SACRG-1, Severe Accident Control Room Guideline Initial Response, because core damage is imminent.

C. Return to Step 4 of FR-S.1 to conti nue efforts to emergency borate the RCS and check for sources of positive reactivity.

D. Transition to SACRG-1, Severe Accident Control Room Guideline Initial Response, to continue efforts to emergency borate the RCS and check for sources of positive reactivity.

Answer: B Explanation:

A. Incorrect. Core damage is imminent, thus SACRG procedures are needed rather than FRGs. B. Correct. Correct action to take IAW Step 14 of FR-S.1.

C. Incorrect. This is incorrect with RCS temperature rising. This option would be correct if temperature was less than 1200°F, per Step 15. D. Incorrect. Correct transition to be made but wrong reason given. Temperature is rising and core damage is imminent thus more severe actions than boration and looking for positive reactivity needs to be done.

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

): FR-S.1 and BD-FR-S.1, Response To Nuclear Power Generation/ATWS

References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-29, Obj A, Explain the Purpose and Major Action Categories of: 1. FR-S.1, Response to Nuclear Power Generation/ATWS.

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: 55.43.5 Comments: SRO level question based on the applicant having to assess the given plant conditions and selecting the correct transition based on the given conditions and the basis for the transition.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 055 - Station Blackout Group # 1 K/A # 055 EA2.03 Importance Rating 3.9 Ability to determine or interpret the following as they apply to a Station Blackout: Actions necessary to restore power.

Question #78 Given the following plant conditions:

• The plant is refueling with fuel movement in progress

• 'B' ESF Emergency Bus, NB02, is de-energized for bus cleaning.

• A loss of offsite power occurs

• 'A' Emergency Diesel Gener ator (EDG), NE01, failed to automatically start and cannot be started from the Control Room

• The Power Supply Supervisor informed the Shift Manager that offsite power will not be available for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Which one of the following proc edures will be implemented, including the preferred action to restore power, by t he Control Room Supervisor?

A. ECA-0.0, Loss of All AC Power; c onnect NB01 to the Alternate Emergency Power Supply (AEPS)

B. ECA-0.0, Loss of All AC Power; attempt to locally start 'A' Emergency Diesel Generator, NE01

C. OTO-EJ-00001, Loss of RHR Flow; attempt to locally start 'A' Emergency Diesel Generator, NE01

D. OTO-EJ-00001, Loss of RHR Flow; c onnect NB01 to the Alternate Emergency Power Supply (AEPS)

Answer: C Explanation:

A. Incorrect. Entry conditions are met for ECA-0.0, but this procedure is only applicable in Modes 1-4 and the plant is in Mode 6. AEPS can be used to restore power to NB01, but it is not the preferred source.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator B. Incorrect. Entry conditions are met for ECA-0.0, but this procedure is only applicable in Modes 1-4 and the plant is in Mode 6. C. Correct. OTO-EJ-00001 is the correct procedure to enter since the plant is in Mode 6 (ECA-0.0 not applicable in given Mode). EDG is preferred source of power for the NB bus. AEPS would be used if the EDG cannot be started. D. Incorrect. Procedure implementation is correct but AEPS is not the preferred source to restore power to NB01. It would only be used if the EDG cannot be locally started.

Technical Reference(s

): Users Guide For Emergency Response Guidelines and Background Documents OTO-EJ-00001, Loss of RHR Flow OOA-ZZ-SSM01, Shutdown Safety Monitoring (SSM)

References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-01, Obj Y, Explain how initial plant conditions affect the applicability of the EOP's.

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: 55.43.5 Comments: SRO level question based on the applicant having to assess the given plant conditions and selecting the correct procedure to restore power to the ESF bus in order to restore shutdown cooling.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 057 - Loss of Vital AC Instrument Bus Group # 1 K/A # 057 2.2.37 Importance Rating 4.6 Ability to determine operability and/or availability of safety related equipment.

Question #79 Given the following plant conditions:

• Reactor power is 100%

• Annun 25B, NN11 INV TRBL/XFR, alarms

• Computer Point NNU0001A, 1E INV NN 11 XFER TO ALT SPLY, is in alarm

• The Secondary Operations Technician reports the following local indications:

• NK01 voltage is 135 volts

• NN01 voltage is 120 volts

• P202, Bypass Source Supplying Load, light is LIT Which one of the following descr ibes the status of the 120 VAC electrical distribution system?

Inverter NN11 is ______(1)________. Instrum ent Bus NN01 is ______(2)________.

(1) (2) A. INOPERABLE INOPERABLE B. OPERABLE INOPERABLE C. OPERABLE OPERABLE D. INOPERABLE OPERABLE

Answer: D Explanation:

A. Incorrect. NN01 is operable as it has correct breaker alignment and voltage. B. Incorrect. NN11 is inoperable as it is connected to its alternate power supply rather than to NK01. NN01 is operable as it has correct breaker alignment and voltage.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator C. Incorrect. NN11 is inoperable as it is connected to its alternate power supply rather than to NK01. D. Correct. NN11 is inoperable as it is connected to its alternate power supply rather than to NK01. NN01 is operable as it has correct breaker alignment and voltage.

Technical Reference(s

): Tech Spec 3.8.7, Electrical Power Systems, Inverters - Operating References to be provided to applicants during examination:

None Learning Objective:

T61.110, LP 6, Obj G, Explain the bases for the Safeguards Power System Technical Specifications (SRO and STA) / Explain the Technical Specifications for the Safeguards Power System.

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: 55.43.2 Comments: SRO level question based on the applicant having to know and apply Tech Specs to determine operability.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 1 W/E11 - Loss of Emergency Coolant Recirculation Group # 1 K/A # W/E11 EA2.2 Importance Rating 4.2 Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation): Adherence to appropriate procedures and operation within the limitations in the facility's license and amendments.

Question #80 Given the following plant conditions:

• The plant was operating at 100% when a Loss of Coolant Accident occurred

• The crew has transitioned to EC A-1.1, Loss of Emergency Coolant Recirculation, due to a loss of both RHR Trains

• While performing actions IAW ECA-1.1, the STA r eports that the Containment Critical Safety Function (CSF) has turned RED due to rising Containment pressure

• All other CSFs are verified to be no higher than YELLOW Which one of the following proc edures would be used by the Control Room Supervisor, per the ERG Rules of Usage, and what action should be taken in response to the High Containment Pressure?

A. Transition to FR-Z.1, Response To Hi gh Containment Pressure, and direct the operators to operate both C ontainment Spray Pumps as directed by FR-Z.1.

B. Transition to FR-Z.1, Response To Hi gh Containment Pressure, and direct the operators to operate both C ontainment Spray Pumps as described by ECA-1.1.

C. Remain in ECA-1.1.

Operate all Containment Spray Pumps until Containment pressure is < 17 psig, then stop and start Containment Spray Pumps as required to maintain pressure < 17 psig.

D. Remain in ECA-1.1. Do not run t he Containment Spray Pumps due to a loss of emergency coolant recirculation capability unless Containment pressure reaches 48 psig.

Answer: B

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

A. Incorrect. Plausible as the transition to FR-Z.1 is correct; however, with the given conditions the containment spray pumps will be operated IAW directions given in ECA-1.1. B. Correct. FR-Z.1 would be entered as directed by the CSFs. Containment spray pumps would be operated as directed by Step 1 RNO per ECA-1.1 while staying in FR-Z.1. C. Incorrect. Plausible if thought that remaining in ECA-1.1 is correct; however, all containment spray pumps would be operated based on RWST level, containment pressure and containment coolers in service.

D. Incorrect. Plausible as direction for operating containment spray pumps is correct; however, with the given conditions a transition would be made to FR-Z.1 rather than remaining in ECA-1.1.

Technical Reference(s

): ECA-1.1, Loss Of Emergency Coolant Recirculation, and FR-Z.1, Response To High Containment Pressure

References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-30, Obj B, Explain the difference between operations of the containment spray system in FR-Z.1 as compared to ECA-1.1. T61.003D, LP D-13, Obj H, Outline procedural flowpath including major system and equipment operation in accomplishing the goal of ECA-1.1, Loss of Emergency Coolant Recirculation.

Question Source: Bank # __X____ Modified Bank # ______ New _______

Question History: Last NRC Exam ___Diablo Canyon Exam_________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.43.5 Comments: SRO level question based on the applicant having to recall what action is required per the EOP and assessing plant conditions and then prescribing what procedure to implement to mitigate and recover the unit.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 077 - Generator Voltage and Electric Grid Disturbances Group # 1 K/A # 077 2.2.36 Importance Rating 4.2 Ability to analyze the effect of maintenance activities, such as degraded power sources, on the status of limiting conditions for operations.

Question #81 Reactor power is 100%.

The Transmission Operations Supervisor has notified the Control Room that a system Category 8 Alarm is present due to maintenanc e activities occurring at the Montgomery substation.

The Transmission Operations Supervisor r eports that the Predicted Contingency Voltage is 319.5 kV.

The following is provided from OSP-NE-00003, Technical Specifications Actions - A.C.

Sources: NB01 and NB02 Powered from separate ESF transformers (Dual Source) or from the same ESF transformer (Single Source) Configuration Required Voltage (kV) Predicted Contingency Voltage InitialsDual Source Single Source LTC(s) AUTO & Cap Banks 372.6 to 329.8 372.6 to 332.9 LTC(s) MAN

• & Cap Banks 372.6 to 335.7 372.6 to 344.3 LTC(s) AUTO & No Cap Banks 372.6 to 341.2 372.6 to 344.3 LTC(s) MAN* & No Cap Banks 372.6 to 347.1

• If in MAN, LTC secondary shall be set (FIXED) at the 13 tap.

Which one of the following actions is requi red to be taken by the Control Room Supervisor based on the information provided?

A. Declare both offsite sources INOPERABLE; Commence an immediate plant shutdown in accordance with LCO 3.0.3.

B. Declare NB01 and NB02 INOPERABLE; Restore ei ther NB01 or NB02 to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator C. Declare both offsite sources INOPER ABLE; Restore one offsite source to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

D. Declare NB01 and NB02 IN OPERABLE; Commence an i mmediate plant shutdown in accordance with LCO 3.0.3.

Answer: C Explanation:

A. Incorrect. Declaring both offsite sources is correct, but action is for two inoperable NB buses rather than two inoperable offsite sources. B. Incorrect. Plausible as the OSP refers to NB bus voltage but if the predicted contingency falls outside the required range, the offsite sources are the component declared inoperable. C. Correct. Actions are correct per TS 3.8.1. D. Incorrect. Plausible as the OSP refers to NB bus voltage but if the predicted contingency falls outside the required range, the offsite sources are the component declared inoperable.

Technical Reference(s

): OSP-NE-00003 Technical Specifications Actions - A.C. Sources, and TS 3.8.1, Electrical Power Systems - AC Sources - Operating

References to be provided to applicants during examination:

None Learning Objective:

T61.0110, LP-6, Obj G, Explain the bases for the Safeguards Power System Technical Specifications (SRO and STA) / Explain the Technical Specifications for the Safeguards Power System.

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: 55.43.2 Comments: SRO level question based on the applicant having to analyze given conditions for Tech Spec applicability and determine required actions that are greater than one (1) hour.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 1 005 - Inoperable/Stuck Control Rod Group # 2 K/A # 005 AA2.03 Importance Rating 4.4 Ability to determine and interpret the following as they apply to the Inoperable / Stuck Control Rod: Required actions if more than one rod is stuck or inoperable.

Question #82 Given the following plant conditions:

• Reactor Power is 100%

• The crew is performing OSP-SF-00002, Control Rod Partial Movement

• During testing, Shutdown (S/D) Bank 'A' Rods D2 and B12 fail to move

• Current rod positions on SF-074, Rod Pos Indication (DRPI), for S/D Bank 'A' Rods D2 and B12 is 228 steps

• All other S/D Bank 'A' rods indicate 210 steps on SF-074

• S/D Bank Step Counters A1 and A2 indicate 212 steps What action must be taken due to the given indications in accordance with Technical Specifications?

A. Restore rods to within alignment limits within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> or be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> B. Verify Shutdown Margin to be within the limits provided in the COLR within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and be in Mode 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

C. Re-evaluate safety analyses and confi rm results remain valid for duration of operation under these conditions

D. Verify Shutdown Margin to be within the limits provided in the COLR within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and reduce thermal power to < 75% RTP within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> Answer: B Explanation:

A. Incorrect. Plausible as alignment is allowed for 1 rod; not allowed for more than 1 rod. B. Correct. Correct action to take IAW TS 3.1.4 Action D.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator C. Incorrect. Plausible as this is allowed for 1 rod if reduced power operation is maintained; maintaining reactor critical would require a re-evaluation of safety analyses. D. Incorrect. Plausible as this is action provided in TS 3.1.4 but not for the specific conditions given.

Technical Reference(s

): Technical Specifications 3.1, Reactivity Control Systems, 3.1.4 Rod Group Alignment Limits

References to be provided to applicants during examination:

None Learning Objective:

T61.0110, LP-26, Obj U, State the Technical Specification limiting conditions for operations (LCOs) applicable to the rod control system.

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: 55.43.2 Comments: SRO level based on the applicant having to recall Tech Spec actions greater than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 1 067 - Plant Fire on Site Group # 2 K/A # 067 2.1.6 Importance Rating 4.8 Ability to manage the control room crew during plant transients.

Question #83 The plant is operating at 100% when the Sh ift Manager enters OT O-ZZ-00001, Control Room Inaccessibility, due to heat and smoke in the Control Room.

The Control Room Supervisor (CRS) will direct the Control R oom Operators to

_______(1)_________ befor e they leave the Control Room to go to the

_________(2)_________.

A. (1) trip the main turbine (2) Auxiliary Shutdown Panel

B. (1) trip all reactor coolant pumps (2) Auxiliary Shutdown Panel

C. (1) trip the main turbine (2) Auxiliary Feedwater Pump rooms

D. (1) trip all reactor coolant pumps (2) Auxiliary Feedwater Pump rooms

Answer: D Explanation:

A. Incorrect. Turbine is not tripped. SM initially goes to the ASP, not the CRS. B. Incorrect. SM initially goes to the ASP, not the CRS. C. Incorrect. Turbine is not tripped. D. Correct. Per OTO-ZZ-00001, the CRS will direct the control room operators to trip all RCPs. The CRS will be assigned Attachment D when he leaves the CR and will go the AFW pump rooms to perform Attachment D.

Technical Reference(s

): OTO-ZZ-00001, Control Room Inaccessibility References to be provided to applicants during examination:

None Learning Objective: T61.0003B, LP 31, Obj E, Given a set of plant conditions or parameters NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator indicating Control Room Inaccessibility, Identify the correct procedure(s) to be utilized and Outline the high level actions to stabilize the plant.

Question Source: Bank # ______ Modified Bank # ______ New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.43.5 Comments: SRO level question based on the applicant having to assess given plant conditions and select procedure action to mitigate the given conditions and know specific actions required of the SRO position.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 069 - Loss of Containment Integrity Group # 2 K/A # 069 2.4.41 Importance Rating 4.6 Knowledge of the emergency action level thresholds and classifications.

Question #84 Given the following plant conditions:

• Containment pressure is 38 psig

• Containment temperature is 220°F

• NB01 is deenergized due to a bus lockout

• The RCS is superheated

• Containment Radiation Monitors GT RE-59 and 60 are reading 1.3E5 R/hr

• Dose assessments indicate dose at the Site Boundary is 150 mRem TEDE Which one of the following describes:

1) The Emergency Action Level (EAL) required to be declared

AND

2) The Protective Action Recommendations (PARs) that should be issued for this EAL?

(Refer to EIP-ZZ-00101, Addendum 1, which is provided for reference)

A. 1) General Emergency 2) Evacuate ALL people within a 2 mile radius around the plant AND 5 miles downwind of the plant in affected sectors.

B. 1) General Emergency 2) Shelter ALL people within a 2 mile radius around the plant AND 5 miles downwind of the plant in affected sectors.

C. 1) Site Area Emergency 2) Shelter ALL people within a 2 mile radius around the plant AND 5 miles downwind of the plant in affected sectors.

D. 1) Site Area Emergency 2) Evacuate ALL people within a 2 mile radius around the plant AND 5 miles NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator downwind of the plant in affected sectors.

Answer: A Explanation:

A. Correct. Classification and PAR are correct IAW EIP-ZZ-00101 and EIP-ZZ-00212. B. Incorrect. Classification is correct but PAR is incorrect. PAR for a GE defaults to evacuation rather than shelter. C. Incorrect. Incorrect as both classification and PAR are incorrect.

D. Incorrect. Incorrect classification; PAR is correct.

Technical Reference(s

): EIP-ZZ-00101, Addendum 1, Emergency Action Level Classification Matrix, and EIP-ZZ-00212, Protective Action Recommendations References to be provided to applicants during examination: EIP-ZZ-00101, Addendum 1

Learning Objective: T68.1020, LP 1, Obj B, Determine the emergency classification for given indications and/or symptoms, per EIP-ZZ-00101, and Obj H, State the protective actions that must be recommended at a General Emergency, per EIP-ZZ-00212.

Question Source: Bank # __X____ Modified Bank # ______

New _______

Question History: Last NRC Exam ___Turkey Point 2011 Exam_________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.43.5 Comments: SRO level question based on the applicant having to declare the proper EAL and PAR.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 1 W/E01 - Rediagnosis Group # 2 K/A # W/E01 EA2.1 Importance Rating 4.0 Ability to determine and interpret the following as they apply to the (Reactor Trip or Safety Injection Rediagnosis): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Question #85 ES-0.0, Rediagnosis, has been entered following a reactor trip and safety injection.

The following conditions are noted by the control room crew:

• Containment pressure 6.5 psig

• Containment temperature 165°F

• Steam Generator level 40%

WR and lowering in all SGs

• Steam Generator pressure 650 psig and lowering in all SGs

• Pressurizer level 0%

• RCS pressure 1680 psig

• MSIVs All closed Which one of the following describes t he proper transition to be made from ES-0.0 to respond to the above plant conditions?

Transition to-

A. E-1, Loss of Reactor or Secondary Coolant B. E-2, Faulted Steam Generator Isolation C. ECA-1.2, LOCA Outside Containment D. ECA-2.1, Uncontrolled Depressu rization of All Steam Generators Answer: D Explanation:

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator A. Incorrect. Plausible as this is a transition that can be made from ES-0.0. Given conditions include some indication of a RCS leak. B. Incorrect. Plausible as this could be a transition from ES-0.0; however, with the given conditions, ECA-2.1 is transitioned to rather than E-2.

C. Incorrect. Plausible as this is a transition that can be made from ES-0.0. Given conditions include some indication of a RCS leak. D. Correct. Transition is correct IAW ES-0.0, Step 1 RNO.

Technical Reference(s

): ES-0.0, Rediagnosis References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-5, Obj D, State and Explain the parameters which are evaluated, including their Criteria and Basis, to transition from ES-0.0 to other procedures.

Question Source: Bank # __003D050J01A (L2472)____ Modified Bank # ______

New _______

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.43.5 Comments: SRO level question based on the applicant having to assess plant conditions and then prescribing a procedure to mitigate the given plant conditions.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 012 - Reactor Protection System Group # 1 K/A # 012 A2.03 Importance Rating 3.7 Ability to (a) predict the impacts of the following malfunctions or operations on the RPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Incorrect channel bypassing.

Question #86 Given the following plant conditions:

• Reactor power is 100%

• Pressurizer Pressure Chann el, BB PT-457, fails low

• The crew enters OTO-BB-00006, Pressurizer Pressure Control Malfunction, to respond to the failure and stabilize the plant

• When tripping bistables for Channel 457 IAW OTO-BB-00006, bistables for Channel BB PT-456 are inadver tently tripped due to improper communication

(1) What Emergency Operation Pr ocedure will be implemented after E-0, Reactor Trip or Safety Injection, is complete?

AND

(2) What is the reporta bility for this event?

A. (1) ES-0.1, Reactor Trip Response (2) Reportable as this is a valid signal

B. (1) ES-0.1, Reactor Trip Response (2) Not reportable as this is not a valid signal

C. (1) ES-1.1, SI Termination (2) Reportable as this is a valid signal

D. (1) ES-1.1, SI Termination (2) Not reportable as this is not a valid signal

Answer: C

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

A. Incorrect. Plausible as this response for procedure implementation would be correct if operator does not recognize that an SI has occurred. Also, reportability requirement is correct. B. Incorrect. Plausible as this response for procedure implementation would be correct if operator does not recognize that an SI has occurred. Reportability portion is not correct. C. Correct. An SI will occur since two PZR pressure channels will have low bistables tripped. Any ESF signal is reportable. D. Incorrect. Plausible response as procedure implementation is correct. Incorrect due to second part concerning reportability. Even if op erator induced an event is reportable.

Technical Reference(s

): E-0,Reactor Trip or Safety Injection APA-ZZ-00520, Reporting Requirements And Responsibilities Dwg 7250D64, Sheet 6, Pressurizer Trip Signals

References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-4, Obj L, Outline procedural flowpath including major system and equipment operation in accomplishing the goal of E-0. T61.0110, LP 69, Obj D.2, Perform the following as they pertain to APA-ZZ-00520, "Reporting Requirements and Responsibilities": 2. Discuss the incidents reportable in the following time frames: e. Immediate (4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />).

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: 55.43.1 and 5 Comments: SRO level question based on the applicant having to assess given plant conditions and then prescribe a procedure flowpath to mitigate and recover the plant to a stable condition. Also SRO knowledge required for reportability criteria.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 2 061 - Auxiliary / Emergency Feedwater System Group # 1 K/A # 061 2.2.22 Importance Rating 4.7 Knowledge of limiting conditions for operations and safety limits.

Question #87 Given the following plant conditions:

• Reactor power is 50%

• The Turbine Driven AFW pump was taken OOS 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> ago for bearing replacement

• Both Motor Driven AFW pump motors have been determined to be INOPERABLE due to the discovery of Foreign Material in their oil bubblers Which one of the following actions is required?

A. Restore at least 1 AFW train to service within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or be in Hot Standby within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

B. Restore at least 2 AFW trains to servic e within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, or be in Hot Standby within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />

C. Enter TS 3.0.3; be in at least Hot St andby within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br /> from the time it was determined that no AFW trains` were operable

D. Initiate action to restore 1 AFW train to operable status immediately; when 1 AFW train is restored, be in Hot Standby within the following 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> Answer: D Explanation:

A. Incorrect. No shutdown is required for this condition as there would be no AFW available as a source of feed during the shutdown. B. Incorrect. Plausible as this would be correct with 2 AFW trains inoperable; however there are three inoperable trains.

C. Incorrect. Plausible as ESW suction lost to two AFW trains at 0700 but the TDAFW train does NOT become inoperable due to having two ESW supply headers. D. Correct. Per TS 3.7.5, with no operable AFW trains, immediately initiate action to restore one NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator train to operable status; 6 hour6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> clock to Mode 3 starts when 1 train is restored to operable.

Note: AFW is a unique system concerning TS 3.0.3 and having no system components operable and its impact on other plant systems. TS 3.0.3 is not applicable and Mode changes or power reductions are suspended until one AFW train is restored to an operable status because it could force the unit into a less safe condition.

Technical Reference(s

): TS 3.7.5 and TS Bases 3.7.5, Plant Systems - Auxiliary Feedwater (AFW) System

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 25, Obj G, State the Tech Spec LCO for the AFW system and Condensate Storage Tank and Describe what is needed for system operability.

Question Source: Bank # __X____ Modified Bank # ______ New _______

Question History: Last NRC Exam ___2008 Harris___

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.43.2 Comments: SRO level question based on the applicant having to know required TS actions and associated basis.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 2 063 - D.C. Electrical Distribution Group # 1 K/A # 063 2.1.20 Importance Rating 4.6 Ability to interpret and execute procedure steps.

Question #88 Given the following plant conditions:

• Reactor power is 6%

• 'A' Emergency Diesel Generat or (EDG), NE01, is in service, connected to the grid

• Annun 25B, NN11 INV TRBL/XFR, alarms

• Annun 25C, NK01 TROUBLE, alarms

• NK II-I, BATTERY CHG NK21 AMPS 0 amps

• NK II-2, BATTERY NKII AMPS DISCHARGE CHARGE 0 amps

• NK EI-I, 125V DC BUS NK01 VOLT 0 volts (1) Which one of the following procedures should be implemented by the Control Room Supervisor?

AND

(2) What actions will be directed to mitigate the condition?

A. (1) OTO-NK-00001, Failure of NK Battery Charger (2) Place Swing Charger NK25 in service

B. (1) OTO-NK-00002, Loss of Vital 125 VDC Bus (2) Place Swing Charger NK25 in service

C. (1) OTO-NK-00002, Loss of Vital 125 VDC Bus (2) Close the NE01 fuel racks until EDG stops

D. (1) OTO-NK-00001, Failure of NK Battery Charger (2) Close the NE01 fuel racks until EDG stops

Answer: C

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

A. Incorrect. Given plant conditions indicate a loss of a DC bus, not the loss of a charger. If a charger was lost, then there would be battery amps indicated. Direction for placing swing charger would only be applicable if there was a loss of a charger. B. Incorrect. Given plant conditions indicate a loss of a DC bus, not the loss of a charger. If a charger was lost, then there would be battery amps indicated. Direction for placing swing charger would only be applicable if there was a loss of a charger. C. Correct. Given plant conditions indicate a loss of a DC bus, not the loss of a charger. If a charger was lost, then there would be battery amps indicated. With a loss of DC, the EDG cannot be secured and the OTO provides direction to close the fuel racks at the EDG. D. Incorrect. Given plant conditions indicate a loss of a DC bus, not the loss of a charger. If a charger was lost, then there would be battery amps indicated.

Technical Reference(s

): OTA-RK-00016, Addendum 25C, NK01 Trouble OTA-RK-00016, Addendum 25B, NN11 Inverter Trouble/Transfer OTO-NK-00002, Loss of Vital 125 VDC Bus

References to be provided to applicants during examination:

None Learning Objective: T61.0110, LP 6, Obj A, Draw and Explain a one line diagram of the Safeguards Power System to include the components and subsystems listed in objective B with normal breaker alignments shown, and Obj B, Describe the purpose and operation of the following Safeguards Power System components and subsystems: 5. 125 VDC System (NK). T61.003B, LP B-26, Obj C., Given a set of plant conditions or parameters indicating a Failure of an NK Battery Charger, Identify the correct procedure(s) to be utilized and Outline the high level actions to stabilize the plant.

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: 55.43.5 Comments: SRO level question based on the applicant having to assess plant conditions for an abnormal condition and select an abnormal procedure to respond to and mitigate the given condition, including required actions.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 2 064 - Emergency Diesel Generators Group # 1 K/A # 064 A2.06 Importance Rating 3.3 Ability to (a) predict the impacts of the following malfunctions or operations on the ED/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Operating unloaded, lightly loaded, and highly loaded time limit.

Question #89 Given the following plant conditions:

• The unit is in Mode 4

• 'A' Emergency Diesel Gener ator (EDG), NE01, is in a post maintenance run

• It has run for 26 hours3.009259e-4 days <br />0.00722 hours <br />4.298942e-5 weeks <br />9.893e-6 months <br /> at 1.15 MW

• Maintenance has requested that t he Control Room secure the D/G Which one of the following describes:

(1) the impact on NE01?

AND

(2) the required action?

A. 1) Oil will build up in the exhaust 2) Lower load on NE01 to 0.2 MW and secure the EDG IAW OTN-NE-0001A, Standby Diesel Generat ion System - Train A B. 1) Lube oil tem perature will be low 2) Lower load on NE01 to 0.2 MW and secure the EDG IAW OTN-NE-0001A, Standby Diesel Generat ion System - Train A C. 1) Oil will build up in the exhaust 2) Raise load to > 50% for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> IAW OTN-NE-0001A, Standby Diesel Generation System - Trai n A, then secure NE01 D. 1) Lube oil tem perature will be low 2) Raise load to > 50% for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> IAW OTN-NE-0001A, Standby Diesel Generation System - Trai n A, then secure NE01

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

A. Incorrect. Correct impact but action is incorrect for running a D/G for an extended time period at low load IAW OTN-NE-0001A. B. Incorrect. Incorrect impact and action is incorrect for running a D/G for an extended time period at low load IAW OTN-NE-0001A. C. Correct. Correct response IAW OTN-NE-0001A.

D. Incorrect. Action is correct for extended period of low load operation of NE01 but the impact is incorrect.

Technical Reference(s

): OTN-NE-0001A, Standby Diesel Generation System - Train A

References to be provided to applicants during examination:

None Learning Objective:

T61.0110, LP 3, Obj N, Explain the precautions, limitations and bases for the following components/processes associated with OTN-NE-0001A/OTN-NE-0001B, "Standby Diesel Generation System - Train A/B": 3. Extended low load operation.

Question Source: Bank # __X____ Modified Bank # ______ New _______

Question History: Last NRC Exam __Wolf Creek 2011 Exam__________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.43.5 Comments: SRO level question based on the applicant having to assess given plant conditions and then select a procedure to mitigate, including required actions.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 2 103 - Containment System Group # 1 K/A # 103 A2.03 Importance Rating 3.8 Ability to (a) predict the impacts of the following malfunctions or operations on the containment system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Phase A and B isolation.

Question #90 Given the following plant conditions:

• Annun 59B, CISB, alarms

• The Reactor Operator reports the following containment pressure indications:

• GN PI-934 1.0

• GN PI-935 1.5

• GN PI-936 1.0

• GN PI-937 2.0 Which of the following procedures:

(1) Will be implemented by the C ontrol Room Supervisor (CRS)?

AND

(2) What action must be taken by the CRS to satisfy Tech Spec 3.6.3, Containment Isolation Valves?

(1) Procedure (2) CRS Action

A. OTO-SA-00001, Engineered Safety Feature Actuation Verification And

Restoration Locally remove power from and open Containment Isolation CCW Bypass Valves to supply Component Cooling

Water B. OTO-SA-00002, Spur ious Containment Spray And Containment Phase B

Isolation Recovery Place a dedicated op erator locally at Containment Isolation CCW Bypass

Valves opened to supply Component Cooling Water C. OTO-SA-00002, Spur ious Containment Spray And Containment Phase B Locally remove power from and open Containment Isolation CCW Bypass NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Isolation Recovery Valves to supply Component Cooling Water D. OTO-SA-00001, Engineered Safety Feature Actuation Verification And

Restoration Place a dedicated op erator locally at Containment Isolation CCW Bypass

Valves opened to supply Component Cooling Water

Answer: B Explanation:

A. Incorrect. OTO-SA-00001 would not be entered for a spurious CISB actuation. It would be entered if the actuation was valid.

B. Correct. Administrative control taken IAW Step 7 of OTO-SA-00002 and allowed per TS 3.6.3. C. Incorrect. Plausible action as this would allow the Control Room to restore CCW to the containment but it is not allowed by procedure or Tech Spec.

D. Incorrect. OTO-SA-00001 would not be entered for a spurious CISB actuation. It would be entered if the actuation was valid.

Technical Reference(s

): OTO-SA-00002, Spurious Containment Spray And Containment Phase B Isolation Recovery Tech Specs 3.6.3, Containment Isolation Valves ODP-ZZ-00002, Equipment Status Control

References to be provided to applicants during examination:

None Learning Objective: T61.003B, LP B-54, Obj B, Describe symptoms or entry conditions for a Spurious Containment Spray and Containment Phase B Isolation Recovery, and Obj D, Given a set of plant conditions or parameters indicating a required spurious Containment Spray and Containment Phase B Isolation Recovery, 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 ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.43.5 Comments: SRO level question based on the applicant having to assess given plant conditions and then select the correct procedure and recall what strategy or action is written into the procedure, including when the strategy or action is required, to mitigate the given abnormal conditions and NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator satisfy Technical Specifications.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 2 034 - Fuel Handling Group # 2 K/A # 034 K4.02 Importance Rating 3.3 Knowledge of design feature(s) and/or interlocks(s) which provide for the following: Fuel movement.

Question #91 Given the following plant conditions:

• The Plant is in a refueling outage

• Core offload is in progress

• An adjustment needs to be made to the Refuel Machine bridge upender zone interlock that requires the Ref uel Machine to be placed into Bypass Operation

Which one of the following correctly identifies the individual s who may approve placing the Refuel Machine into Bypass Operation?

A. Refueling Reactor Engineer AND Shift Manager

B. Refueling SRO AND Director Nuclear Operations

C. Refueling Reactor Engineer AND Director Nuclear Operations D. Refueling SRO AND Shift Manager

Answer: D A. Incorrect. Plausible as the Refueling Reactor Engineer and Refueling SRO oversee fuel movement during offload activities.

B. Incorrect. DNO not authorized as he is not currently licensed. C. Incorrect. Plausible as the Refueling Reactor Engineer and Refueling SRO oversee fuel movement during offload activities and DNO can authorize most activities that the SM has responsibility for. D. Correct. Per OTS-KE-00013 to place the refuel machine in bypass requires approval of the Refueling SRO and a second SRO.

Technical Reference(s): OTS-KE-00013, Refueling Machine References to be provided to applicants during examination:

None NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Learning Objective: T61.003E 6, LP-5, Obj E.5, Describe the function, capacity, and operation of the following equipment: Refueling machine.

Question Source: Bank # __X____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____Callaway 2011 Reexam________

Question Cognitive Level: Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.43.7 Comments:

Comments: SRO level question based on applicant having to know Refuel floor SRO responsibilities associated with the requirements for operating fuel handling equipment. ROs have no responsibilities for any fuel handling associated equipment or procedures outside of the Control Room.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 2 041 - Steam Dump System and Turbine Bypass Control Group # 2 K/A # 041 2.1.23 Importance Rating 4.4 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Question #92 Given the following plant conditions:

• Reactor was tripped from 100% power

• All four Reactor Coolant Pumps were tripped due to a loss of Component Cooling Water

Which of the following describes:

(1) What direction is provided by the CRS to mitigate the above given conditions?

AND

(2) What is the reason for this action?

A. (1) Transfer Condenser Steam Dumps to Steam Pressure Mode (2) Prevent RCS Overcooling

B. (1) Fast Close Main Steam Isolation Valves (2) Maintain Steam Generator Inventory

C. (1) Transfer Condenser Steam Dumps to Steam Pressure Mode (2) Maintain Steam Generator Inventory

D. (1) Fast Close Main Steam Isolation Valves (2) Prevent RCS Overcooling Answer: A Explanation:

A. Correct. Direction is correct IAW ES-0.1 and reason is correct IAW BD-ES-0.1. B. Incorrect. Plausible as SG pressure will be lowering until steam dumps are placed in steam pressure mode due to excessive RCS cooling and the criteria for closing MSIVs is lowering SG NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator pressure. Reason is also plausible as stopping steam flow from a SG will help to maintain inventory. C. Incorrect. Correct direction but reason is not correct. Reason is plausible as inventory will be maintained if the steam flow is lowered from a SG; however that is not a concern for the given conditions. D. Incorrect. Plausible as SG pressure will be lowering until steam dumps are placed in steam pressure mode due to excessive RCS cooling and the criteria for closing MSIVs is lowering SG pressure.

Technical Reference(s

): ES-0.1 and BD-ES-0.1, Reactor Trip Response

References to be provided to applicants during examination:

None Learning Objective: T61.003D, LP D-6, Obj C, Describe the requirements and basis for the Continuous Action Steps of ES-0.1, Reactor Trip Response, and Obj H, Outline procedural flowpath including major system and equipment operation in accomplishing the goal of ES-0.1.

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: 55.43.5 Comments: SRO level question based on the applicant having to assess given plant conditions and recalling action written into Emergency Operating Procedures and when to implement the strategy to mitigate the given plant conditions and also to recall the basis for the action to be taken.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 2 045 - Main Turbine Generator System Group # 2 K/A # 045 A2.17 Importance Rating 2.9 Ability to (a) predict the impacts of the followin g malfunctions or operation on the MT/G system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Malfunction of electrohydraulic control.

Question #93 Given the following plant conditions:

• Reactor power is 90%

• While conducting OSP-AC-00003, Turbine Control Valve Stroke Test, during the performance of Step 6.1.3, the Operator presses and holds the CV-2 TEST pushbutton

• TURB AUTO STOP 63 AST 2 light is LIT on SB069, Rx Trip/Bypass Permissive Status Panel

• Upon releasing the CV-2 TEST pushbutton the test solenoid fails to de-energize

The CV is (1) and the Control Room Supervisor would direct the Balance of Plant operator to (2) to stabilize the plant.

A. (1) closed (2) lower load using the standby load set potentiometer

B. (1) closed (2) lower load using the load set potentiometer

C. (1) open (2) lower load using the load set potentiometer

D. (1) open (2) lower load using the standby load set potentiometer

Answer: B Explanation:

A. Incorrect. The standby unit would only be used if the load set potentiometer does not work. There is no information provided that indicates the load set potentiometer will not work. B. Correct. The test solenoid stays energized which keeps the CV closed. Action is correct IAW OTO-MA-00001, Turbine Load Rejection.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator C. Incorrect. Valve remains closed due to the failed test solenoid valve; action is correct. D. Incorrect. The standby unit would only be used if the load set potentiometer does not work. There is no information provided that indicates the load set potentiometer will not work.

Technical Reference(s

): OSP-AC-00003, Turbine Control Valve Stroke Test, and OTO-MA-00001, Turbine Load Rejection References to be provided to applicants during examination:

None Learning Objective: T61.003B, LP B-44, Obj B, Describe symptoms or entry conditions for OTO-MA-00001, Turbine Load Rejection, and Obj D, Given a set of plant conditions or parameters indicating a Turbine Load Rejection, Analyze the correct procedure(s) to be utilized and the required actions to stabilize the plant.

Question Source: Bank # __X____ Modified Bank # ______ New _______

Question History: Last NRC Exam ____Wolf Creek 2011 NRC Exam________

Question Cognitive Level: Memory or Fundamental Knowledge _____ Comprehension or Analysis __X___

10 CFR Part 55 Content: 55.43.5 Comments: SRO level question based on the applicant having to assess given plant conditions and direct operator action based on correct procedure implementation to mitigate the given conditions.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 3 GENERIC Group # 1 K/A # 2.1.5 Importance Rating 3.9 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Question #94 A waiver of work hour limitations is permitted ONLY when necessary

______(1)_______.

AND

The _____(2)_______ can approve the requested waiver IAW APA-ZZ-00905, Limitations of Callaway Pl ant Staff Working Hours.

A. (1) to mitigate a condi tion adverse to safety.

(2) Shift Manager (SM)

B. (1) to mitigate a condi tion adverse to safety. (2) Emergency Duty Officer (EDO)

C. (1) due to the declaration of an emer gency, as defined by the Radiological Emergency Response Plan.

(2) Shift Manager (SM)

D. (1) due to the declaration of an emer gency, as defined by the Radiological Emergency Response Plan.

(2) Emergency Duty Officer (EDO)

Answer: A Explanation:

A. Correct. Per APA-ZZ-00905, Section 4.8 and 4.9. B. Incorrect. Plausible as most approvals for waivers, etc., require either EDO or SM approval. C Incorrect. Plausible as emergency declarations are made to mitigate or prevent safety situations; however, support for an emergency declaration is specifically excluded from needing approval to exceed working hour limitations.

D. Incorrect. Plausible as emergency declarations are made to mitigate or prevent safety NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator situations; however, support for an emergency declaration is specifically excluded from needing approval to exceed working hour limitations.

Technical Reference(s

): APA-ZZ-00905, Limitations of Callaway Plant Staff Working Hours References to be provided to applicants during examination:

None Learning Objective:

T61.003A, LP A-19, Obj A, Discuss the following as they pertain to APA-ZZ-00905, "Limitations of Callaway Plant Staff Working Hours": 1. The maximum permitted working hours 2. Who authorizes deviations to working hour restrictions

Question Source: Bank # ______ Modified Bank # ______ New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.43.1 Comments: SRO level question based on applicant having to know administrative controls and approvals for shift staffing requirements.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 3 GENERIC Group # 1 K/A # 2.1.23 Importance Rating 4.4 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Question #95 As the Control Room Supervisor, whic h one of the following Off-Normal procedures would you implement in parallel with E-0, Reactor Trip or Safety Injection?

A. OTO-NK-00002, Loss of Vital 125 VDC Bus B. OTO-ZZ-00008, Steam

/Feedwater Line Break C. OTO-BG-00001, Pressurizer Level Control Malfunction D. OTO-SA-00001, Engineered Safety Feature Actuation Verification And Restoration Answer: A Explanation:

A. Correct. Note at beginning of procedure provides directions that procedure should be performed in parallel with E-0. B. Incorrect. Procedure purpose states procedure is to be used when a reactor trip or safety injection is not required. C. Incorrect. No direction is given to use procedure in parallel with E-0. D. Incorrect. Step 1 in procedure states to go to E-0 if a reactor trip has occurred.

Technical Reference(s

): OTO-NK-00002, Loss Of Vital 125 VDC Bus ODP-ZZ-00025, EOP/OTO User's Guide 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 # ______ New ___X____

Question History: Last NRC Exam ____________

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Question Cognitive Level: Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.43.5 Comments: SRO level question based on applicant having to have knowledge of procedure implementation in coordination with Emergency Operation Procedures above just knowing entry conditions.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 3 GENERIC Group # 2 K/A # 2.2.7 Importance Rating 3.6 Knowledge of the process for conducting special or infrequent tests.

Question #96 You have been assigned as the Test Lead fo r OSP-BB-VL006, RCS Pressure Isolation Valves Inservice Tests-IPTE, for the upcoming Refuel.

ALL of the following items ar e REQUIRED to be covered during the IPTE Job Briefing EXCEPT:

A. Potential risks and High Consequence Activities B. Procedure timeline starting with current plant conditions C. Termination criteria, contingen cy actions and restart criteria D. Contacts, with phone numbers, for Test Lead, SM and Engineering

Answer: D Explanation:

A. Incorrect. Required per APA-ZZ-0100A, Step 4.3.3.

B. Incorrect. Required per APA-ZZ-0100A, Step 4.3.3. C. Incorrect. Required per APA-ZZ-0100A, Step 4.3.3. D. Correct. Plausible as contact information is always good to have; however, during an IPTE evolution, personnel associated with the activity that would be needed will either be located locally at the activity or in the Control Room. This information is not required for the Pre-Job Brief.

Technical Reference(s

): APA-ZZ-0100A, Infrequently Performed Test Or Evolution Guidance References to be provided to applicants during examination:

None Learning Objective:

T61.003A, LP A-14, Obj E, Discus s APA-ZZ-0100a, Infreq uently Performed Test Or Evolution to include the following: 1. Define Infrequently Performed Tests or Evolutions (IPTE) 2. Infrequently Performed Test or Evolution Procedure Development NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator

3. Preparations and Performance of Infrequently Performed Tests or Evolutions Question Source: Bank # ______ Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.43.5 Comments: SRO level question based on applicant having to have the knowledge of administrative requirements concerning the conduct of Infrequently Performed Test or Evolution (IPTE) which could affect the Margin of Safety for the plant.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 GENERIC Group # 2 K/A # 2.2.17 Importance Rating 3.8 Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator.

Question #97 PAL02, Turbine Driven Auxiliary Feedwater Pump, is scheduled for maintenance.

Which one of the following plant equipment is ALLOWED to be taken out of service simultaneously with PAL02?

A. EPGUB3051, Security Diesel Generator B. EDGPA5001, AEPS Diesel Generator #1 C. PAP01, Non-Safety Auxiliary Feedwater Pump D. EPGUB7001, TSC Emer gency Diesel Generator

Answer: D Explanation:

A. Incorrect. Not allowed per ODP-ZZ-00002, APP 02. B. Incorrect. Not allowed per ODP-ZZ-00002, APP 02.

C. Incorrect. Not allowed per ODP-ZZ-00002, APP 02. D. Correct. No restrictions associated with the TSC EDG with a PAL02 outage.

Technical Reference(s

): ODP-ZZ-00002 APP 2, Risk Management Actions For Planned Risk Significant Activities

References to be provided to applicants during examination:

None Learning Objective:

Question Source: Bank # ______ Modified Bank # ______ New ___X____

Question History: Last NRC Exam ____________

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis _____

10 CFR Part 55 Content: 55.43.5 Comments: SRO level question based on applicant having to have the knowledge of the work control process including maintenance prioritization to minimize plant risk, which is an SRO responsibility.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 3 GENERIC Group # 3 K/A # 2.3.14 Importance Rating 3.8 Knowledge of radiation or contamination hazards that may arise during normal, abnormal, or emergency conditions or activities.

Question #98 Given the following plant conditions:

• A large break Loss of C oolant Accident has occurred

• A General Emergency has been declared

• A repair team was sent into Containm ent for emergency repairs on a valve to stop the release

• Due to an accident, both repair team members have been injured, one with a life threatening head injury

(1) What is the maximum allowed dose t hat can be authorized for a rescue team in accordance with HDP-ZZ-01450, Authorization To Exceed Federal Occupational Dose?

AND

(2) Who can approve this dose?

(1) (2) A. 100 rem DDE Radiat ion Protection Manager B. 10 rem DDE Radiation Protection Manager C. 10 rem DDE Recovery Manager D. 100 rem DDE Recovery Manager

Answer: D

Explanation: A. Incorrect. RPM cannot approve.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator B. Incorrect. RPM cannot approve; correct dose is 100 rem, not 10 rem. C. Incorrect. Correct dose is 100 rem, not 10 rem. D. Correct. Per Step 3.3 and Attachment 1 of HDP-ZZ-01450.

Technical Reference(s): HDP-ZZ-01450, Authorization To Exceed Federal Occupational Dose 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. 2. Discuss the limits for plant emergencies and the selection criteria associated with these limits.

Question Source: Bank # _______ Modified Bank # _______

New ___X____

Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___ Comprehension or Analysis ____

10 CFR Part 55 Content:

55.43.4 Comments: SRO level question based on applicant having to assess plant conditions and then prescribe procedure actions and authorizations to mitigate the abnormal event for actions required of an SRO.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 3 GENERIC Group # 4 K/A # 2.4.9 Importance Rating 4.2 Knowledge of low power/shutdown implications in accident (e.g, loss of coolant accident or loss of residual heat removal) mitigation strategies.

Question #99 Given the following plant conditions:

• The plant is in Mode 4

• SI accumulators are isolated

• Pressurizer level is lowe ring in an uncontrolled manner

Which one of the following proce dures will the Control Room Supervisor use to mitigate this event?

A. OTO-BB-00010, Shutdown LOCA B. OTO-EJ-00001, Loss of RHR Flow C. E-0, Reactor Trip or Safety Injection D. OTO-EJ-00003, Loss of RHR While Operating at Reduced Inventory or Mid-Loop Conditions

Answer: A Explanation:

A. Correct. OTO-BB-00003, Step 2 RNO directs the CRS to OTO-BB-00010 for the given indications. OTO-BB-00010 can also be entered directly based on the given conditions. B. Incorrect. Plausible as OTO-EJ-00001 would be transitioned to if the plant was in Mode 5 or 6. C. Incorrect. Plausible as E-0 is usually entered to address RCS leaks. Mode of applicability for E-0 to be used is Mode 1, 2 or 3 with SI accumulators aligned for injection. D. Incorrect. Plausible as OTO-EJ-00003 is used for loss of RCS inventory but not at the given conditions, Technical Reference(s

): OTO-BB-00003, RCS Excessive Leakage OTO-BB-00010, Shutdown LOCA References to be provided to applicants during examination:

None NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Learning Objective: T61.003E, LP E-8, Obj B, Describe symptoms or entry conditions for OTO-BB-00010, Shutdown LOCA, 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: 55.43.5 Comments: SRO level question based on applicant having to assess plant conditions and select appropriate procedure to mitigate given conditions.

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 1 Tier # 3 GENERIC Group # 4 K/A # 2.2.28 Importance Rating 4.1 Knowledge of procedures relating to a security event (non-safeguards information).

Question #100 Which one of the following conditions would require the Control Room Supervisor to direct a reactor trip?

A. NRC Notification of an electrical grid threat B. A Safe Shutdown Earthquake is confirmed C. A Probable Airborne Threat has been received D. A CODE RED plant announcement by Security

Answer: D Explanation:

A. Incorrect. No requirement in OTO-SK-00004 to perform a reactor trip; only requirement is to monitor specific plant indicators and equipment. B. Incorrect. SSE does not require a trip just becau se of the earthquake. A trip could be required or even occur if there was damage.

C. Incorrect. Trip would only be required if airborne threat was imminent. D. Correct. Actions are correct IAW OTO-SK-00001.

Technical Reference(s

): OTO-SK-00001, Plant Security Event - Hostile Intrusion OTO-SK-00002, Plant Security Event - Aircraft Threat OTO-SK-00004, Plant Security Event - Electrical Threat OTA-RK-00024, Addendum 98B, SSE

References to be provided to applicants during examination:

None Learning Objective: T61.003B, LP B-16, Obj D, Given a Plant Security Event - Hostile Intrusion, Identify the correct procedure(s) to be utilized and Outline the high level actions to stabilize the plant.

Question Source: Bank # ______ Modified Bank # ______

New ___X____

NRC Site-Specific Written Examination Callaway Plant Senior Reactor Operator Question History: Last NRC Exam ____________

Question Cognitive Level: Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

10 CFR Part 55 Content: 55.43.5 Comments:

CW-2013-03-Initial License Written Exam RO Reference Handouts

1. Q 2, Q43 - Steam Tables

2. Q18 - Curve Book Figure 10

-1 thru 4, Minimum Excitation Limits at Indicated Terminal Voltages SRO Reference Handout

1. Q84 - EIP-ZZ-00101, Addendum 1