ML20276A154

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Pressurizer Vapor Space Accident Group # 1 K/A # 00008 AK3.05 Importance Rating 4.0 Knowledge of the reasons for the following responses as they apply to the Pressurizer Vapor Space Accident: ECCS termination or throttling criteria.

Question #2 A reactor trip and safety injection have occurred. The crew is in E-1, Loss of Reactor or Secondary Coolant, at Step #6 evaluating the following plant conditions to determine if Safety Injection (SI) termination criteria is met.

RCS pressure is 1300 psig and stable

RCS temperature is 540°F and lowering

Core Exit TC (CETCs) are 555°F and slowly lowering

PZR level is 90% and rising

Containment radiation is 100 R/hr and rising

Containment Pressure is 2 psig and slowly rising

Total AFW flow is 250,000 lbm/hr

SG NR levels are 15%

What describes whether or not SI flow can be terminated and the reason why or why not?

A. SI termination criteria is met because CETCs are lowering B. SI termination criteria is NOT met because RCS pressure is not rising C. SI termination criteria is NOT met because RCS subcooling is less than the required value D. SI termination criteria is met because core cooling is assured due to the presence of a secondary heat sink Answer: C Explanation:

The above plant parameters model a PZR vapor space accident from 100% reactor power.

Note Saturation temperature at 1300 psig is ~579°F.

NRC Written Examination Callaway Plant Reactor Operator E-1 step #6 states "CHECK If ECCS Flow Should Be Reduced"

a. RCS subcooling - GREATER THAN 30°F [50°F]

b. Secondary heat sink:

Narrow range level in at least one intact SG - GREATER THAN 7% [25%]

OR

Total feed flow to intact SGs - GREATER THAN 270,000 LBM/HR

c. RCS pressure - STABLE OR RISING

d. PZR level - GREATER THAN 9% [29%]

e. Go To ES-1.1, SI Termination, Step 1 If the plant conditions are not met by ANY of the above criteria, the RNO is applicable which states "GO TO Step #7" and SI termination is not allow at that time.

A. Incorrect - both the decision and the reason are wrong. SI termination is not allowed due to another parameter (subcooling) not being greater than its minimum value. Core Exit thermal couples are used to calculate subcooling margin and hence it is plausible that they are checked for SI termination criteria and the fact that Core exit thermalcouple tempatures are lowering making it a plausible reason. Additionally, CETCs are used during the performance of E-3, SGTR, step #6 "Initiate RCS Cooldown" making it easy to incorrectly apply them to E-1 step #6.

B. Incorrect - While the decision not to terminate SI is correct, the reason is wrong. RCS pressure can either be stable or rising in order to reduce SI flow and this explanation implies that RCS pressure must be going up in order to allow SI termination which is NOT correct.

C. Correct -SI termination is not allowed because RCS subcooling is below min required value.

Total amount of subcooling is 579°F - 555°F = 24F, which is below the minimum criteria for subcooling, 30°F.

D. Incorrect - While there is a secondary heat sink as SG NR levels are greater than the minimum of 7% (no adverse containment) making it a plausible choice, SI termination is not allowed due to another parameter (subcooling) not being greater than its minimum value.

Technical Reference(s):

1. E-1, Loss of Reactor or Secondary Coolant, Rev 20 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP #8, Objective A: EXPLAIN the Purpose and Major Action Categories of E-1, Loss of Reactor or Secondary Coolant.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3____

10 CFR Part 55 Content:

Comments: To balance the question 2 met and 2 not met choices were presented.

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Large Break LOCA Group # 1 K/A # 000011 EA1.05 Importance Rating 4.3 Ability to operate and monitor the following as they apply to a Large Break LOCA: Manual and/or automatic transfer of suction of charging pumps to borated source Question #3 The Unit was at 100% power with the "A" RHR Pump Out of Service, when it experienced a Large Break LOCA with an automatic Safety Injection.

The crew has just completed ES-1.3, Transfer To Cold Leg Recirculation Alignment.

How many Charging pumps are running and what is the suction source for the running pump(s)?

Charging pump(s) running Suction source A. 1 B RHR pump discharge B. 1 directly from the recirc sump C. 2 B RHR pump discharge D. 2 directly from the recirc sump Answer: C Explanation: Per ES 1.3 EM8807A/B and EJ8804A/B are open. 'A' RHR pump feeds CCPs and SIPs thru EJ8804A and EM8807A/B while 'B' RHR pump feeds CCPs and SIPs thru EJ8804B and EM8807A/B. Even though 'A' RHR pump is OOS, the 'B' RHR pump will feed both CCPs.

SIPs and CCPs do not have suction directly from containment recirc sump.

A. Incorrect - would have 2 CCPs running from B RHR pump discharge B. Incorrect - would have 2 CCPs running from B RHR pump discharge not directly from recirc sump C. Correct - 2 CCPs running from B RHR pump discharge D. Incorrect - would have 2 CCPs running from B RHR pump discharge not directly from recirc sump Technical Reference(s):

1. ES-1.3 Transfer to Cold Leg Recirculation, Rev 013

NRC Written Examination Callaway Plant Reactor Operator

2. M-22EJ01, P&ID RHR System, Rev 062

3. M-22EM01, P&ID SI System, Rev 038

4. M-22BG03, P&ID CVCS system - sheet 3, Rev 059 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP#11, Objective G, OUTLINE procedural flowpath including major system and equipment operation in accomplishing the goal of ES-1.3, Transfer to Cold Leg Recirculation.

Question Source: Bank # ______

Modified Bank # ______

New ____X___

Question History: Last NRC Exam ____NA______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Reactor Coolant Pump Malfunctions Group # 1 K/A # 00015 AK1.02 Importance Rating 3.7 Knowledge of the operational implications of the following concepts as they apply to Reactor Coolant Pump Malfunctions (Loss of RC Flow): Consequences of an RCPS failure Question #4 Reactor Power is 50%.

'A' RCP No. 1 Seal & Bearing Inlet Temperature is 180°F and rising at 1.0°F/min.

'A' RCP Shaft Vibration is 9 mils and rising at 0.1 mils/min.

(1) What 'A' RCP parameter requires the crew to take action First?

And (2) Per OTO-BB-00002, RCP Off-Normal, what is/are the correct action(s)?

A. (1) Shaft Vibration (2) Trip the affected RCP ONLY B. (1) Shaft Vibration (2) Trip the reactor and then trip the affected RCP C. (1) No. 1 Seal & Bearing Inlet Temperature (2) Trip the affected RCP ONLY D. (1) No. 1 Seal & Bearing Inlet Temperature (2) Trip the reactor and then trip the affected RCP Answer: D Explanation:

Based on the values and rates given in the stem, the No. 1 Seal & Bearing Inlet temp will reach its actionable values of 230F in 50 minutes. (230 - 180F) / 1.0 F/min = 50 The Shaft vibration component has 2 actionable limits:

ALL RCPs vibration on the shaft - LESS THAN 20 MILS (Step A1)

ALL RCPs vibration on the shaft - LESS THAN 15 MILS but rises at a rate greater than 2 Mils/hr. (Step A2 and RNO)

NRC Written Examination Callaway Plant Reactor Operator It would take 110 minutes to reach the actionable values of Step A1. (20 - 9 mils) / .1 mil/min It would take 60 minutes to reach the actions of Step A2 RNO. (15 - 9 mils) / 0.1 mil/min = 60 minutes.

Therefore, based on the conditions provided No. 1 Seal & Bearing Inlet Temperature will be the first actionable limit reached.

Step B4 states "CHECK No. 1 Seal & Bearing Inlet Temperature - LESS THAN 230°F ON ALL RCPs" and the RNO applies after the calculated time is reached which states "IF Reactor power is greater than or equal to 48% (P-8 lit), THEN PERFORM the following:

Step A2 for RCP vibrations states ALL RCPs vibration on the shaft - LESS THAN 15 MILS and when it exceeds 15 mils at the 60 minute point the RNO should apply which states

a. IF vibration on the frame rises at a rate of greater than 1 Mil/hr OR vibration on the shaft rises at a rate greater than 2 Mils/hr, THEN PERFORM ONE of the following:

IF Reactor power is greater than or equal to 48% (P-8 lit), THEN Go To Attachment D, RCP AND Reactor Trip.

Tripping the RCP only is plausible if the power level is not correctly applied, i.e. P-8 setpoint is not correctly remembered.

A. Incorrect - See above explanation - both are wrong B. Incorrect - See above explanation - wrong parameter C. Incorrect - See above explanation - wrong action D. Correct Technical Reference(s):

1. OTO-BB-00002, RCP Off-Normal, Rev 34 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off normal Operations, LP#11; Objective D & F:

D: Given a set of plant conditions or parameters indicating a RCP Off-Normal condition, ANALYZE the correct procedure(s) to be utilized and the required actions to stabilize the plant.

F. EXPLAIN the vibration levels at which action must be taken and DESCRIBE the required actions.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

NRC Written Examination Callaway Plant Reactor Operator Comprehension or Analysis __X___

LOD ____3____

10 CFR Part 55 Content:

Comments:

k/a match as the applicant must apply RCP system knowledge and select the required plant components to trip which is a consequence of the RCP failure

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Loss of Reactor Coolant Makeup Group # 1 K/A # 000022 AK3.02 Importance Rating 3.5 Knowledge of the reasons for the following responses as they apply to the Loss of Reactor Coolant Makeup: Actions contained in SOPs and EOPs for RCPs, loss of makeup, loss of charging, and abnormal charging Question #5 With the unit at full power, events result in a loss of ALL seal injection to the RCPs.

OTO-BB-00002, RCP Off-Normal, has been implemented.

With the above conditions, the BASES for maintaining cooling flow to the RCP thermal barrier heat exchanger is to A. minimize CVCS system contamination B. provide adequate seal cooling to the RCPs C. minimize flashing in the seal injection lines D. prevent steam bubble collapse in the CCW system Answer: B Explanation: When seal cooling is lost, thermal barrier Hx will provide RCP seal cooling as RCS flows past Hx and into #1 seal leak off line.

A. Incorrect, RCS would flow past thermal barrier Hx and flow to VCT, bypassing CVCS demin and raising CVCS activity.

B. Correct, see explanation above.

C. Incorrect, Check valve in seal injection line just upstream of RCP will prevent backflow into seal injection line and prevent flashing.

D. Incorrect, steam intrusion into CCW system would not occur as long as CCW remains aligned to thermal barrier Hx, while a positive side affect of maintaining CCW flow it is not the bases.

Technical Reference(s):

1. OTO-BB-00002, RCP Off-Normal, Rev 034

2. M-22BG01, P&ID CVCS System - Sheet 1, Rev 033

3. M-22BG03, P&ID CVCS System - Sheet 3, Rev 059

NRC Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination: None Learning Objective: T61.003B, Off-Normal Operations, LP #11, Objective C: DESCRIBE Continuous Action Step(s) including the required Response Not Obtained actions.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___NA_________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

LOD ____3______

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Loss of Residual Heat Removal System Group # 1 K/A # 000025 AK2.01 Importance Rating 2.9 Knowledge of the interrelations between the Loss of Residual Heat Removal System and the following: RHR heat exchangers Question #6 The plant is in Mode 4 performing a cooldown to Mode 5:

RHR trains A and B are in service

A tube leaks develops in the Train A RHR Heat Exchanger What is the effect of the tube leak and the indications available to the operator to determine if leakage exists?

EFFECT Fluid leaks from INDICATIONS A. RHR to CCW Surge tank level ONLY B. CCW to RHR Surge tank level ONLY C. RHR to CCW Surge tank level AND Radiation Monitoring D. CCW to RHR Surge tank level AND Excore NIs Answer: C Explanation:

A: Incorrect: Leakage direction is correct. Surge tank level is not the only indication available.

The CCW system also has a process rad monitor which the student must know exists.

B: Incorrect: Leakage is in the opposite direction, but numerous CCW loads are at a lower pressure than CCW so the student must have knowledge of relative pressures under these plant conditions. Surge tank level is not the only indication available. If this were to occur as stated, a dilution of the RCS would also be taking place.

C: Correct: RHR is at a higher pressure than CCW. Both surge tank level and the CCW rad monitor would be available to diagnose the leak.

D: Incorrect: Leakage is in the opposite direction, but numerous CCW loads are at a lower pressure than CCW so the student must have knowledge of relative pressures under these plant conditions. If this were to occur, both indications would help the operator to diagnose the leakage.

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

1. OTO-EG-00001, CCW System Malfunction, Rev 018 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#7, Objective D: LIST the systems that interface with the RHR System and EXPLAIN how a loss of the interfacing system or a loss of the RHR System affects the other.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

LOD _____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Loss of Component Cooling Water Group # 1 K/A # 000026 G2.1.23 Importance Rating 4.3 Ability to perform specific system and integrated plant procedures during all modes of plant operation.

Question #7 Reactor Power is 40%.

At 0858, CCW has been lost to the A RCP motor.

At 0900, A RCP Motor Parameters are as follows:

o Motor Bearing Temperatures are 190°F and rising at 1°F/min.

o Motor Stator Winding Temperatures are 302°F and rising at 3°F/min.

(1) If the current trends continue, the FIRST 'A' RCP motor component to reach its temperature limit will be the Motor ......?

And (2) What is the LATEST time the 'A' RCP MUST be secured?

A. (1) Bearings (2) 0903 B. (1) Bearings (2) 0905 C. (1) Stator Windings (2) 0903 D. (1) Stator Windings (2) 0905 Answer: C Explanation: Per OTO-EG-00001, if CCW flow is lost to the RCPs, then the crew is directed to OTO-BB-00002 Specifically Attachment C for a loss of CCW to the RCP. The limits in Attachment C are as follows:

1. Motor Bearing Temperatures - less than 195F

2. Motor Stator Winding temperature - less than 311F

3. Time lost - less than 10 minutes

NRC Written Examination Callaway Plant Reactor Operator And if either one of these is exceeded, C1 RNO applies which directs securing the affected RCP.

With the values and trends given, bearing temperatures can rise for 5 minutes, (195-190F)/

1F/min, before the criteria to secure the RCP is met. Motor winding temperatures can rise for 3 minutes, (311-302F) / 3F/min, before the criteria to secure the RCP is met. Neither time exceeds the 10 minutes from the time of CCW lost to the RCPs (i.e. 0908 is not the limiting time to secure the 'A' RCP). Therefore, the RCP must be secured by 0903 to prevent damage to Stator Winding temperature trends.

A. Incorrect - wrong component B. Incorrect - both are wrong C. Correct - see above explanation D. Incorrect - wrong time Technical Reference(s):

1. OTO-EG-00001, CCW System Malfunction, Rev 18

2. OTO-BB-00002, RCP Off-Normal, Rev 34 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP #17, Objective C & E:

C. DETERMINE the affect that a CCW System malfunction has on a cooled system/component and SELECT the subsequent action to respond to the associated malfunction.

E. Given a set of plant conditions or parameters indicating a CCW System Malfunction, ANALYZE the correct procedure(s) to be utilized and the required actions to stabilize the plant.

Question Source: Bank # _X_(no id)____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____2016_Question #6______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

LOD _____3_____

10 CFR Part 55 Content:

10 CFR: 55.41(b)(10)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Pressurizer Pressure Control System Group # 1 Malfunction K/A # 000027AK3.03 Importance Rating 3.7 Knowledge of the reasons for the following responses as they apply to the Pressurizer Pressure Control Malfunctions: Actions contained in EOP for PZR PCS malfunction Question #8 Reactor Power is 100%.

A failure of the controlling input to the Pressurizer Pressure Master Controller, BB PK-455A, caused actual pressurizer pressure to increase to 2285 psig.

The Pressurizer Pressure Master Controller has been placed in MANUAL.

What action is required to return pressure to 2235 psig?

A. Raise the controller output B. Lower the controller output C. Raise the pressure setpoint adjustment D. Lower the pressure setpoint adjustment Answer: B Explanation In manual, the PZR master controller output must be adjusted. If it was in Automatic, the pressure setpoint would be adjusted.

With the data given in the stem, pressure input into the controller must have failed low causing the controller to believe the PZR pressure is low and turn on heater. This would raise PZR pressure past the desired setpoint and per OTO-BB-00006 RNO actions when pressure is out of band, the master controller is placed in MANUAL. The action from there simply say, to return PZR Pressure to in band. To Lower must lower output to open spray (reference OTN-BB-00005 Attachment #1or OOA-BB-00002)

A. Incorrect - plausible as based on the conditions given (desired pressure is lower than the actual pressure that is necessary to raise the output to lower the pressure B. Correct C. Incorrect - wrong action because the controller is in manual D. Incorrect - wrong action because the controller is in manual

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

1. OTN-BB-00005, PZR and Pressurizer Pressure Control, Rev 16

2. OTO-BB-00006, PZR Pressure Control Malfunction, Rev 20 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP #41, Objective C: Given a set of plant conditions or parameters indicating a Pressurizer Pressure Control Malfunction, ANALYZE the correct procedure(s) to be utilized and the required actions to stabilize the plant.

Question Source: Bank # ___X__L16387_____

Modified Bank # ______

New _______

Question History: Last NRC Exam __N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis _X____

LOD ____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 ATWS Group # 1 K/A # 000029 EA1.14 Importance Rating 4.2 Ability to operate and monitor the following as they apply to a ATWS: Driving of control rods into the core Question #9 A Reactor Trip Signal was generated following a Loss of Offsite Power.

The Reactor failed to trip and the CRS has directed the performance of immediate actions in FR-S.1.

The Reactor Operator is inserting Control Bank rods using manual rod control.

At what speed (spm), are the rods inserting?

A. 8 B. 48 C. 64 D. 72 Answer: B Explanation:

A. Incorrect: 8 spm is the minimum rod speed in automatic B. Correct 48 spm is the rod speed in manual for Control Banks C. Incorrect 64 spm is the rod speed in manual for Shutdown Banks D. Incorrect 72 spm is the maximum rod speed in automatic Technical Reference(s):

1. FR-S.1, RESPONSE TO NUCLEAR POWER GENERATION/ATWS, Rev 14

2. Tech Manual M-763-0083, IM FULL LENGTH ROD CONTROL SYSTEM, Rev 13 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP# 26 (Rod Control), Enabling Objective C, DESCRIBE the rod speed program in manual and automatic.

NRC Written Examination Callaway Plant Reactor Operator Question Source: Bank # ____

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___N/A___

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD _____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Steam Generator Tube Rupture Group # 1 K/A # 000038 EA2.16 Importance Rating 4.2 Ability to determine or interpret the following as they apply to a SGTR: Actions to be taken if S/G goes solid and water enters steam line REFERENCE PROVIDED Question #10 The crew is responding to a Steam Generator Tube Rupture and is controlling RCS pressure and charging flow to minimize leakage.

Current plant conditions are:

Pressurizer level 60% and stable

Ruptured SG NR level Offscale High

Containment pressure 0.3 psig Per E-3, Steam Generator Tube Rupture, step 30, what action should be performed?

A. Raise charging flow B. Turn on PZR Heaters C. Lower charging flow or raise letdown flow D. Depressurize RCS below ruptured SG pressure Answer: C

NRC Written Examination Callaway Plant Reactor Operator Distractor B Explanation:

Distractor A Distractor D Correct Answer A. Incorrect plausible for S/G level offscale high but PZR level is less than 25%

B. Incorrect plausible for PZR level of 60% but S/G level is lowering C. Correct for given conditions D. Incorrect plausible for PZR level of 60% but S/G level is rising Technical Reference(s):

1. E-3, Steam Generator Tube Rupture, Rev 024, Table step 30.b References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP#17, Objective O, EXPLAIN Basis for controlling RCS Pressure and Charging flow to minimize RCS to Secondary leakage.

Question Source: Bank # ______

Modified Bank # _X__11959___

New _______

NRC Written Examination Callaway Plant Reactor Operator Question History: Last NRC Exam ___NA______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____2_____

10 CFR Part 55 Content:

Comments:

E-3, Step 30 PZR Ruptured SG(s) NR Level Level Rising Lowering Offscale High RAISE RAISE RAISE LESS THAN charging flow charging flow charging flow 25% [45%]

DEPRESSURIZE RCS MAINTAIN RCS and below ruptured ruptured SG(s)

SG(s) pressure pressures equal using Step 30.c BETWEEN DEPRESSURIZE RCS TURN ON PZR MAINTAIN RCS and 25% [45%] below ruptured heaters: ruptured SG(s)

AND 50% SG(s) pressure pressures equal using Step 30.c Group A PG HIS-19 BB HIS-51A

-OR-Group B BETWEEN DEPRESSURIZE RCS PG HIS-21 MAINTAIN RCS and 50% AND below ruptured BB HIS-52A ruptured SG(s) 74% [64%] SG(s) pressure pressures equal using Step 30.c LOWER Charging LOWER Flow Or RAISE Charging Flow Or Letdown Flow RAISE Letdown Flow GREATER LOWER MAINTAIN RCS and THAN Charging Flow Or ruptured SG(s) 74% [64%] RAISE Letdown pressures equal Flow LOWER Charging Flow Or RAISE Letdown Flow

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Steam Line RuptureExcessive Heat Group # 1 Transfer K/A # 000040 AK1.07 Importance Rating 3.4 Knowledge of the operational implications of the following concepts as they apply to Steam Line Rupture: Effects of feedwater introduction on dry S/G Question #11 What is the effect of adding feedwater to a faulted dry Steam Generator (SG)?

A. Longer RCS cooldown time that causes Pressurized Thermal Shock (PTS)

B. Release to the environment above the 10CFR100 limits on a design basis event C. Higher probability of occurrence of a steam generator tube rupture in the faulted steam generator D. Overfilling faulted steam generator and causing the steam generator atmospheric dump valve to lift Answer: C Explanation:

Per the basis document for CA2.1, step #2 Caution: If feed flow to a SG is isolated and the SG is allowed to dry out, subsequent reinitiation of feed flow to the SG could create significant thermal stress conditions on SG components. Maintaining a minimum verifiable feed flow to the SG allows the components to remain in a "wet" condition, thereby minimizing any thermal shock effects if feed flow is increased.

A. Incorrect - plausible since longer cooldown but no increase in RCS pressure B. Incorrect - this is one of the reasons for E-1 step #16 "If fuel damage has occurred following a LOCA, there is a potential for releasing appreciable quantities of radionuclides from the RCS through pre-existing SG tube leakage, if primary-to-secondary differential pressure is established.

In order to prevent such a release, SG pressures must be maintained greater than RCS pressure when RCS activity is high.

C. Correct - See explanation above D. Incorrect - plausible as in E-3 step #7 check intact SG levels discuss isolating FW flow to prevent overfilling a SG.

Technical Reference(s):

1. ECA-2.1, Uncontrolled Depressurization of all S/Gs, Rev 16

2. BD-ECA-2.1, Basis Document for ECA-2.1, Rev 9

3. ERG Background, item 4-6, HFRH1BG Rev 3, Section 2.4

NRC Written Examination Callaway Plant Reactor Operator

4. BD-E-1, Basis Document for E-1, Rev 16

5. BD-E-3, Basis Document for E-3, Rev 15 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP #16, Objective F, EXPLAIN the Notes and Cautions, including their basis, for ECA-2.1, Uncontrolled Depressurization of all Steam Generators.

Question Source: Bank # ______

Modified Bank # ______

New ___X___

Question History: Last NRC Exam ___NA_______

Question Cognitive Level:

Memory or Fundamental Knowledge ___X___

Comprehension or Analysis _____

LOD _____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Loss of Main Feedwater Group # 1 K/A # 000054 G2.4.4 Importance Rating 4.5 Ability to recognize abnormal indications for system operating parameters that are entry-level conditions for emergency and abnormal operating procedures.

Question #12 Reactor Power is 100% when the following occurs:

Annunciators 108C-111C, SG A-D Lev Dev, alarm

Annunciators 108D-111D, SG A-D Flow Mismatch, alarm

"B" Condensate Pump trips What procedure should the crew enter FIRST?

A. E-0, Reactor Trip or Safety Injection B. OTO-MA-00008, Rapid Load Reduction C. OTO-AE-00001, Feedwater System Malfunction D. OTO-AE-00002, Steam Generator Water Level Control Malfunction Answer: C Explanation:

A. Incorrect - If two condensate pumps trip, the main feedwater system would not be able to supply enough feedwater flow to maintain SG level due to a loss of main feed pump suction pressure.

B. Incorrect: - plausible as power reduction is required which OTO-MA-00008 provides the direction to perform this downpower but wrong as the feedwater system abnormal procedure has the directions included C. Correct: Callaway does not have a annunciator for condensate pump trip. The first indications present would be lowering SG level and a flow mismatch alarm as indicated in the stem. The symptoms or entry conditions for OTO-AE-00001 include:

"Annunciator 108C (109C, 110C, 111C), SG A (B, C, D) Lev Dev

Annunciator 108D (109D, 110D, 111D), SG A (B, C, D) Flow Mismatch" D. Incorrect: plausible since the S/G Level and flow deviations are entry conditions for OTO-AE-00002 but wrong as the cause of the Flow mismatch and low level alarms is the condensate pump trip

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

1. OTO-AE-00001, Feedwater System Malfunction, Rev 39 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off-Normal Operation, LP#8, Objective B, DESCRIBE symptoms or entry conditions for OTO-AE-00001, Feedwater System Malfunction.

Question Source: Bank # ______

Modified Bank # ______

New ___X___

Question History: Last NRC Exam ____NA______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____2______

10 CFR Part 55 Content:

Comments:

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

Question #13 A Loss of All AC has occurred and the crew is performing ECA-0.0, Loss of All AC Power.

What is the approximate battery discharge rate (amps per hour) of NK01 if it will take 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months for the battery to fully discharge? Initially, NK01 was fully charged.

A. 81 B. 109 C. 150 D. 218 Answer: C Explanation:

Per OTO-NB-00001 Step #26 the following are the rates of the given battery

NK01, (1650 amp hours)

NK03, (900 amp hours)

PK01, (2400 amp hours local)

PK03, (2400 amp hours local) 1650 / 11 hours1.273148e-4 days 0.00306 hours 1.818783e-5 weeks 4.1855e-6 months = 150 amps per hour 900 / 11 = 81.8 amps per hours (~81) 2400 / 11 = 218.18 amps per hours (~218)

If it is believed a battery has 1200 amp hour rating, then 1200 / 11 = 109.09 amps per hours

(~109)

A. Incorrect - See above calculations B. Incorrect - See above calculations C. Correct - See above calculations D. Incorrect - See above calculations

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

1. OTO-NB-00001, Loss of Power to NB01, Rev 34 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP #6, Objective M: EXPLAIN the precautions, limitations and bases for the following components/conditions associated with OTN-NK-00001, Class 1E 125 VDC Electrical System:

1. Battery capacity

2. Maximum NK Battery Charge amperage output Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3____

10 CFR Part 55 Content:

(CFR: 41.7 / 45.5 to 45.8)

Comments:

Overlap check with Audit exam Q#47 - no overlap exists Modified original question from a NK01 focus to a PK03 focus

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Loss of Vital AC Instrument Bus Group # 1 K/A # 00057 AA2.19 Importance Rating 4.0 Ability to determine and interpret the following as they apply to Loss of Vital AC Instrument Bus and the following: The plant automatic actions that will occur on the loss of a vital ac electrical instrument bus Question #14 Reactor Power is 100% when the following events occur:

Charging pump suction swaps from the VCT to the RWST

Annunciator 82B, OTT Rod Stop, is LIT

Annunciator 82C, OPT Rod Stop, is LIT

Annunciator 83C, RX Partial Trip, is LIT What event is in progress?

A. Loss of NK01 B. Loss of NN01 C. PR NI 41 is offscale HIGH D. Loop 1 RCS RTD Channel Failure Answer: B Explanation:

Per OTO-NN-00001, Attachment C, the listed automatic actions are indicative of a loss of NN01.

Loss of NK (125 VDC) is plausible as OTO-NK-00002 for a loss of NK01 references PZR level 459 and ensures that that instrument is not selected. Furthermore the NK bus is the normal power supply to the NN bus via an inverter which makes the loss of NK bus plausible also.

Per OTO-BB-00004, annunciator 82B, 82C and 83C are symptoms of a RCS RTD channel failure but there would be no swapover to the RWST making it plausible but incorrect.

A PR NI failing high would cause the annunciators also and it may be believed that it will cause a swapover to the RWST. The SR channel, SR N31, powered from NN01 will cause a swapover to the RWST if it fails high and isnt blocked NOTE: Loss of NN04 would also be a correct answer due to VCT channel 185 and PR NI 44.

NRC Written Examination Callaway Plant Reactor Operator A. Incorrect - See above explanation B. Correct C. Incorrect - See above explanation D. Incorrect - See above explanation Technical Reference(s):

1. OTO-NN-00001, Loss of Safety Related Instrument AC Power, Rev 38

2. OTO-BB-00004, RCS RTD Channel Failures, Rev 21 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP #27, Objective B: DESCRIBE symptoms or entry conditions for OTO-NN-00001, Loss Of Safety Related Instrument Power.

Question Source: Bank # __X_(no id)___

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2019 Q#13_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Loss of Instrument Air Group # 1 K/A # 00065 AK3.03 Importance Rating 2.9 Knowledge of the reasons for the following responses as they apply to the Loss of Instrument Air: Knowing effects on plant operation of isolating certain equipment from instrument air Question #15 The plant is in Mode 3 at Normal Operating Pressure and Temperature (NOP/NOT) with the following conditions:

Annunciator 92A, COMPRESSED AIR PRESS LO, is LIT

KA PI 0040, Instrument Air Hdr Pressure, is 65 psig and lowering (1) OTO-KA-00001, Partial or Total Loss of Instrument Air, directs the crew to close And (2) What is the reason for verifying these valves are closed?

A. (1) PZR PORVs (2) to prevent or minimize radioactive releases from the RCS B. (1) PZR PORVs (2) to prevent a loss of RCS Pressure control and inadvertent safety injection C. (1) PZR Spray Valves (2) to prevent or minimize radioactive releases from the RCS D. (1) PZR Spray Valves (2) to prevent a loss of RCS Pressure control and inadvertent safety injection Answer: D Explanation:

Step #17 of OTO-KA-00001 directs the PZR Spray valves to be closed. While not stated explicitly in OTO-KA-00001, the reason the PZR Spray valves are checked closed is that if they were open PZR pressure would lower to the point of SI upon restoration of air pressure. PZR PORVS are plausible as they on the same system and inside containment. Inadvertent opening of these would cause a RCS pressure loss also. Both the PORVs and the Spray Valve can / will discharge into the PRT and may cause the PRT rupture disc to relieve as in the TMI accident

NRC Written Examination Callaway Plant Reactor Operator making "to prevent or minimize the radioactive releases from the RCS" a plausible distractor.

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

1. OTO-KA-00001, Partial or total Loss of Instrument Air, Rev 29

2. OTA-RK-00024 ADD 92A, Compressed Air Low, Rev 0 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP #20, Objective C: Given a set of plant conditions or parameters indicating a Partial or Total Loss of Instrument Air, IDENTIFY the correct procedure(s) to be utilized and OUTLINE the high level actions to stabilize the plant.

Question Source: Bank # ______

Modified Bank # __X L22387____

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 LOCA Outside Containment / 3 Group # 1 K/A # W/E04 EK2.1 Importance Rating 3.5 Knowledge of the interrelations between the (LOCA Outside Containment) and the following:

Components, and functions of control and safety systems, including instrumentation, signals, interlocks, failure modes, and automatic and manual features.

Question #16 Operators are performing actions of ECA-1.2, LOCA Outside Containment.

(1) What action MUST be done in order to CLOSE EMHV8835, SI Pumps to Cold Leg Injection Valve?

And (2) If during the performance of ECA-1.2, RCS pressure and PZR level begin to rise, the control room crew should enter ____(2)______.

A. (1) Place the Power Lockout Switch in the NON ISO position (2) E-0, Reactor Trip or Safety Injection B. (1) Place the Power Lockout Switch in the NON ISO position (2) E-1, Loss of Reactor or Secondary Coolant C. (1) Open EMHV8802A&B, SI Pump Discharge to Hot Leg Injection Valves (2) E-0, Reactor Trip or Safety Injection D. (1) Open EMHV8802A&B, SI Pump Discharge to Hot Leg Injection Valves (2) E-1, Loss of Reactor or Secondary Coolant Answer: B Explanation:

Per ECA-1.2, step #2b, PLACE the following Power Lockout switches in the NON ISO position" this is required to reposition this valve as motive power needs to be returned to the valve operator by placing the power Lockout switch to the Non ISO (not isolated) position. If left in the ISO position (normal lineup), no motive power will be available for the valve actuator.

EMHV8835 receives an open signal on an SI and should be open when the operators start to perform this step.

The distractor of EMHV8802A&B is plausible because they are in parallel flow paths to the RCS.

NRC Written Examination Callaway Plant Reactor Operator One flow path is to the RCS hot leg and one to the RCS cold leg. With the data given in the stem, i.e there is a LOCA accident in progress, it may be falsely believed that the operator is required to establish an ECCS flowpath into the RCS hot legs before the ECCS flowpath to the RCS cold legs are isolated. Specifically, ECCS flow into the RCS must be available prior to closing a valve i.e. 8802A&B must be open before EMHV8835 is closed as there is a LOCA in progress.

Step #3 of ECA-1.2 directs the operator to either transition to E-1 if the leak outside containment is isolated or to ECA-1.1 if it was NOT successfully isolated. Since the stem asks if the leak is successfully isolated, the correct answer is E-1. ECA-1.1 is plausible if the candidate does not properly recall step #3 of ECA-1.1 (i.e ECA-1.1 is in step #3 RNO) but could be considered a SRO level distractor.

E-0, Reactor Trip or Safety Injection, is plausible as ECA-1-1 can be entered from E-0 at Step

20 and it may be believed that ECA-1.1 includes a step / statement of "return to Procedure step and effect" Per ODP-ZZ-00025, steps 4.10.2.b and 4.10.3 which state: "When branching to another procedure, the CRS should note the current Procedure and Step in effect when this branching occurred. This is required to assist in subsequent branching back to that point if directed." and "Some procedures contain transition worded Return to procedure and step in effect. These transitions can occur from either column. When encountered, the CRS must determine which procedure was being implemented (in effect) immediately prior to the current procedure."

This is incorrect as there is no need to enter E-0 to determine that a transition to E-1 is required.

This would delay emergency response time and mitigation efforts.

A. Incorrect -See above explanation - wrong procedure B. Correct - See above explanation C. Incorrect - See above explanation - both are wrong D. Incorrect - See above explanation - wrong action to close EMHV8835 Technical Reference(s):

1. ECA-1.2, LOCA Outside Containment, Rev 7

2. M-22EM01, P&ID High Pressure Coolant Injection System, Rev 39

3. ODP-ZZ-00025, EOP/OTO User's Guide. Rev 38 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP #17 Safety Injection, Objective B:

DESCRIBE the purpose and operation of the following SI System components:

8. SI to Cold Leg Isolation Valve (EMHV8835)

T61.003D, Emergency Operations, LP #14, ECA-1.2, Objective F: OUTLINE procedural flowpath including major system and equipment operation in accomplishing the goal of ECA-1.2.

T61.003D, Emergency Operations, LP #14, ECA-1.2 Objective E: STATE and EXPLAIN the parameters which are evaluated, including their Criteria and Basis, to transition from ECA-1.2 to other procedures.

Question Source: Bank # ______

NRC Written Examination Callaway Plant Reactor Operator Modified Bank # ___X_(no bank id)__

New _______

Question History: Last NRC Exam ____2017 NRC Q#3 ________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

Comments:

Resetting SI (done in step #2.a) is also a correct answer for part 1)

K/A match as this safety system component that is manipulated during the performance of ECA-1.2 and the candidate must understand the features and interlocks (or lack there off) of this component. Not all ECCS valves have this Power ISO / NON ISO Lockout switch feature.

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Loss of Emergency Coolant Group # 1 Recirculation K/A # W/E11 EK2.2 Importance Rating 3.9 Knowledge of the interrelations between the (Loss of Emergency Coolant Recirculation) and the following: Facilitys heat removal systems, including primary coolant, emergency coolant, the decay heat removal systems, and relations between the proper operation of these systems to the operations of the facility.

Question #17 What describes the correct order of operations required in ECA-1.1, Loss of Emergency Coolant Recirculation, when Annunciator 47A, RWST EMPTY, alarm is received, and RWST level is verified to be 5%?

A. Stop BOTH Containment Spray pumps and reduce ECCS flow to ONE train running B. Stop BOTH Containment Spray pumps, and throttle SI and RHR flow in accordance with decay heat removal requirements C. Stop ALL pumps taking a suction from the RWST, align normal charging, and initiate secondary depressurization to facilitate SI Accumulator injection D. Stop ONE Containment Spray Pump and reduce ECCS flow to ONE train running.

Ensure sufficient Containment Fan Coolers aligned, then stop the second Containment Spray Pump Answer: C Explanation:

Per the continuous action step #6, if RWST is less than 6%, the RNO applies and directs the operator to step #30 which STOP Pumps Taking Suction From RWST And PLACE Switches In PULLTOLOCK Position, step #31 - 33 direct aligning normal charging and secondary depressurization to inject Accumulator as necessary.

A. Incorrect but plausible as Step #7 (table) determines that BOTH Containment spray pumps should be secured and Step #2 RNO directs trying to restore at least one train of ECCS as ECCS may be required for injection. This distractor securing both CTMT spray pumps but leaves one train of ECCS running, which is plausible as described above, but incorrect as the Caution pror to step #2 states "IF suction source is lost to any ECCS or Containment Spray pump(s), the pump(s) should be stopped."

NRC Written Examination Callaway Plant Reactor Operator B. Incorrect - See above explanation for distractor A. Plausible as it may be believed that decay heat RHR and SI flow from the RWST may be allowed in order to maintain current core temperatures.

C. Correct D. Incorrect but plausible as the order of steps provides for some containment cooling before spray is stopped but incorrect as one train of ECCS would still be running aligned from the RWST.

Technical Reference(s):

1. ECA-1.1, Loss of Emergency Coolant Recirculation, Rev 14 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP# 13, Objective C: DESCRIBE the requirements and basis for the Continuous Action Steps of ECA-1.1, Loss of Emergency Coolant Recirculation.

Question Source: Bank # __X__L16516__

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____4____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Loss of Secondary Heat Sink Group # 1 K/A # W/E05 EK1.3 Importance Rating 3.9 Knowledge of the operational implications of the following concepts as they apply to the (Loss of Secondary Heat Sink) Annunciators and conditions indicating signals, and remedial actions associated with the (Loss of Secondary Heat Sink).

Question #18 The crew has entered FR-H.1, Response to Loss of Secondary Heat Sink.

The crew is at step 4 to check NR SG level > 7% when the BOP reports the following:

"A", "B", and "D" SG wide range levels are 25% and lowering slowly

"C" SG wide range level is 30% and stable

Total AFW flow is 100,000 lbm/hr

All SG pressures are 1040 psig

RCS pressure is 1600 psig What action(s) should be taken next?

A. Exit FR-H.1 and Return to Procedure and step in effect B. Trip all RCPs and initiate RCS Bleed and Feed in accordance with FR-H.1 C. Raise the setpoint of "C" SG ASD to conserve SG inventory in accordance with FR-H.1 D. Initiate secondary depressurization to establish Condensate flow in accordance with FR-H.1 Answer: B Explanation:

A. Incorrect - plausible because if AFW flow is > 270,000 lbm/hr this would be the correct answer or as C SG is greater than 25% WR the setpoint of 25% NR could falsely be applied to this condition B. Correct step 2 of FR-H.1 is a continuous action step for when 3 SGs < 27% wide range level C. Incorrect - plausible since raising ASD setpoint would conserve SG water inventory ASD are manipulated during step #9 of FR-H.1 D. Incorrect - plausible if SG levels not recognized as being too low. This is step #9 of FR-H.1.

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

1. FR-H.1, Response to Loss of Secondary Heat Sink, Rev 018 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP#26, Objective 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 # _X L14115____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____2005____

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

LOD _____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Pressurizer Level Control Malfunction Group # 2 K/A # 000028 AK3.03 Importance Rating 3.5 Knowledge of the reasons for the following responses as they apply to the Pressurizer Level Control Malfunctions: False indication of PZR level when PORV or spray valve is open and RCS saturated Question #19 A plant heatup is being performed in accordance with OTG-ZZ-00001, Plant Heatup Cold shutdown to Hot Standby with the following conditions:

RCS temperature is 430°F

RCS pressure is 685 psig

One Pressurizer PORV fails open and does not re-close

As RCS pressure lowers to 325 psig, the operators notice Pressurizer level rising (NOTE: Assume NO manual operator actions have taken place.)

With pressurizer level rising, the liquid that is now entering the Pressurizer originates from ..

A. the RHR Pumps B. the SI Pumps C. the Safety Injection Accumulators D. a void which has formed in the reactor vessel head Answer: D Explanation:

A: INCORRECT: Plausible because it is a possible source of water for the RCS. However, with the given conditions, there would be no reason for the pump to be running (auto SI actuation was not enabled yet).

B: INCORRECT: Plausible because it is a possible source of water for the RCS. However, with the given conditions, there would be no reason for the pump to be running (auto SI actuation was not enabled yet).

C: INCORRECT: Plausible because it is a possible source of water for the RCS. Based on the given conditions, the accumulators have not yet been placed in service (outlet valves would still be closed).

NRC Written Examination Callaway Plant Reactor Operator D: CORRECT: The given pressure/temperature conditions results in a saturated RCS leading to a void in the head and a rise in PZR level.

Technical Reference(s):

1. OTG-ZZ-00001, Plant Heatup Cold shutdown to Hot Standby, Rev 093

2. BD-CSF-1, CSFST Basis, Rev 006 References to be provided to applicants during examination: None Learning Objective:T61.003D, Emergency Operations, LP#3, Objective B & C:

B. DESCRIBE the response of specified plant parameters to the following major accidents.

Small LOCA

Steam break inside and outside containment

Feed break inside and outside containment

Steam generator tube rupture C. DESCRIBE the following specified plant parameters:

Tavg

Pressurizer level

VCT level

Containment pressure, temperature, and humidity

Containment airborne activity

Steam flow

Steam pressure

Feedwater flow

Steam generator level

Condenser/SG blowdown activity Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____NA_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____2_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Loss of Source Range Nuclear Group # 2 Instrumentation K/A # 000032 AK1.01 Importance Rating 2.5 Knowledge of the operational implications of the following concepts as they apply to Loss of Source Range Nuclear Instrumentation: Effects of voltage changes on performance Question #20 The plant is in Mode 6.

Both Source Range detectors are reading approximately the same when the high voltage power supply for SE NI-31 fails LOW.

What type of detector are the Source Range NIs and what is the effect of the high voltage power supply failure on SE NI-31 indication?

A. BF3 proportional counter, NI-31 indicates LESS than NI-32 B. Compensated ion chamber, NI-31 indicates LESS than NI-32 C. BF3 proportional counter, NI-31 indicates the SAME as NI-32 D. Compensated ion chamber, NI-31 indicates the SAME as NI-32 Answer: A Explanation:

A: CORRECT: It is a proportional counter which means the output will be proportional to the voltage causing a lower reading with a lower voltage.

B: INCORRECT: Plausible detector type since other excore NIs are ion chambers C: INCORRECT: Plausible effect because this would be correct for an ion chamber D: INCORRECT: Either answer is plausible based on the discussion above Technical Reference(s):

1. Westinghouse Technical Manual Nuclear Instrumentation, section 6.1 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#28, Objective B, DESCRIBE the purpose and operation of the following Nuclear Instrumentation System (NIS) components: Source Range Nuclear Instruments

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

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____NA________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

LOD ____2______

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Loss of Condenser Vacuum Group # 2 K/A # 00051 AA2.02 Importance Rating 3.9 Ability to determine and interpret the following as they apply to the Loss of Condenser Vacuum:

Conditions requiring reactor and/or turbine trip Question #21 Reactor Power is 30% with main generator connected to grid.

Annunciator 116B, Condenser Vacuum Lo, is LIT

The crew has entered OTO-AD-00001, Loss of Condenser Vacuum

Condenser backpressure is 7 inches HgA and steady

All 3 Circ Water Pumps are running What is the NEXT action required per OTO-AD-00001?

A. Place Rod Control Handswitch, SE HS-9, in Auto B. Manually trip the Reactor and go to E-0, Reactor Trip Or Safety Injection C. Manually trip the Turbine and go to OTO-AC-00001, Turbine Trip Below P-9 D. Check Runback / Setback Status - Vacuum Runback is active and turbine load is lowering and if not manually reduce turbine load Answer: B Explanation:

Notes:

Annunciator 116B, alarms when 2/3 signals on 1/3 exhaust hood reaches 6.5" Hg abs (DEHC mod 2019)

Annunciator 117B will alarm if a setback or runback is active and greater than 40%

intermediate power

Annunciator 118B, Setback / Runback Disabled, alrams when the setback function is disabled manually in DEHC For the conditions given in the stem, the correct procedural flow path is Step #1 to Step #3 and at step #3 neither of the conditions listed are occurring and the RNO is applicable which directs the crew to step #22. At step #22, the RNO applies and as backpressure is not improving and power is greater than 10%, a manually reactor trip and entry into E-0 are directed.

The distractor of "Place Rod Control Handswitch, SE HS-9, in Auto" is plausible as it is step #6

NRC Written Examination Callaway Plant Reactor Operator but incorrect as this step is not performed.

The distractor of "Check Runback / Setback Status - Vacuum Runback is active and turbine load is lowering and if not manually reduce turbine load" is from the RNO of Step #10 and is plausible as vacuum is greater than 5.5 inches HgA but incorrect as this step and its RNO are not performed.

The distractor of "Manually trip the Turbine and go to OTO-AC-00001, Turbine Trip Below P-9" is plausible as it is also in step #22 RNO and applicable when power is less than 10%. Furthermore the given stem power level is less than P-9 (50%) making this a plausible choice.

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

1. OTO-AD-00001, Loss of Condenser Vacuum, Rev 35 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP #7, Objective G: RECOGNIZE the conditions that would require a Reactor Trip / Turbine Trip.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____2____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Area Radiation Monitoring System Group # 2 K/A # 000061 G2.1.30 Importance Rating 4.4 Ability to locate and operate components, including local controls Question #22 While moving fuel assemblies in the Fuel Building, area radiation monitors SD RE-37 and 38 go into HI alarm.

What conditions are necessary for the audible and flashing alarm indications to reset?

A. Both will reset as soon as the alarm is acknowledged B. Both will reset when the alarm is acknowledged only if the alarm condition is clear C. The audible alarm will reset only when the alarm is acknowledged, the flashing alarm will reset only when the condition clears D. The flashing alarm will reset only when the alarm is acknowledged, the audible alarm will reset only when the condition clears Answer: C Explanation: The audible "high alarm" condition may be acknowledged on the Remote indicator, but will not reset until the alarm condition is cleared. The visual flashing alarm light on the Remote Indicator will continue to function until the alarm condition is cleared.

A. Incorrect - See above explanation - visual flashing alarm light on the Remote Indicator will continue to function until the alarm condition is cleared B. Incorrect - See above explanation - the audible can be reset before the condition clears C. Correct D. Incorrect - See above explanation - these reset are opposite of the correct indications as presented in choice C.

Technical Reference(s):

1. E-23SD04, Area Radiation Monitors - Central Processors Power Supply & Alarms, Rev 004 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP-36, Objective B. DESCRIBE the purpose and

NRC Written Examination Callaway Plant Reactor Operator operation of the following Process and Area Radiation Monitoring components:

11. Area Radiation Monitors Question Source: Bank # _X_L16243_

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2005_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3______

10 CFR Part 55 Content:

10 CFR 55.41(b)(11)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 High Reactor Coolant Activity Group # 2 K/A # 00076 AK2.01 Importance Rating 2.6 Knowledge of the interrelations between the High Reactor Coolant Activity and the following:

Process radiation monitors Question #23 With the unit at power, which of the following describes the radiation monitor(s) that will provide direct confirmation of High RCS Activity in accordance with OTO-BB-00005, RCS High Activity?

A. SJ-RE-01, CVCS Letdown Monitor B. GT-RE-59, CTMT High Range Area Monitor C. GT-RE-31, Containment Atmosphere Monitor D. GT-RE-21B, Unit Vent Air Exhaust Radiation Monitor Answer: A Explanation:

Per OTO-BB-00005, SJ RE-01 is a symptom or entry condition.

OTA-SP-RM011 Attachment 19, provides direction for a high GT-RE-59 (CTMT High Range Area Monitor) alarm. This alarm is either indicative of Excess RCS leakage or a fuel handling accident but not high RCS activity with an intact RCS boundary.

OTA-SP-RM011 Attachment 18, provides direction for a high GT-RE-31(CTMT Atmosphere) alarm. GT-RE-31 contains a gas, iodine, and particulate channel and is used during mostly during containment ventilation activities such as purge operations (backup to GT-RE 22 and 33) and serves similar functions that GT-RE-59 provides i.e. Excess RCS leakage or a fuel handling accident.

OTA-SP-RM011 Attachment 16 provides direction for a high GT-RE-21B (Unit Vent Rad Monitor).

This rad monitor detects high rad in the containment ventilation exhaust as it leaves the unit vent and is an indication of an elevated release due to a LOCA in containment.

A. Correct - Per OTO-BB-00005, section B, this is the only radiation monitor that is a symptom or entry condition for this procedure.

B. Incorrect - Plausible as this alarm would be present during LOCA or excessive RCS leakage conditions and is also used in the Fission Product Barrier Matrix for EAL determinations.

NRC Written Examination Callaway Plant Reactor Operator C. Incorrect - Plausible as this alarm would be present during LOCA or excessive RCS leakage conditions D. Incorrect - Plausible as this alarm would be present during LOCA or excessive RCS leakage conditions with containment ventilation elevating the release via the unit vent.

Technical Reference(s):

1. OTO-BB-00005, RCS High Activity, Rev 18

2. OTA-SP-RM011, Rad Monitor Control Panel, RM-11, Rev 44 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off normal Operations, LP #14, OTO-BB-00005, Objective C:

DESCRIBE symptoms or entry conditions for OTO-BB-00005, RCS High Activity.

Question Source: Bank # ___X L16535___

Modified Bank # ______

New _______

Question History: Last NRC Exam ____2016 Q#22________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

10 CFR 55.41(b)(11)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 SI Termination Group # 2 K/A # W/E02 EK1.2 Importance Rating 3.4 Knowledge of the operational implications of the following concepts as they apply to the (SI Termination): Normal, abnormal and emergency operating procedures associated with (SI Termination).

Question #24 Reactor Power was 100% when a small break LOCA occurred.

The crew has transitioned to ES-1.1, SI Termination, and are currently establishing letdown per step #14.

Current plant conditions are:

The MSIVs have just closed due to a slow rise in containment pressure

RCS Subcooling is 45°F and slowly lowering

Pressurizer Level is 20% and slowly lowering

RCS Pressure is 1800 psig and slowly lowering What are the NEXT actions required per ES-1.1?

A. Restart one CCP and go to E.2, Faulted Steam Generator Isolation B. Re-establish ECCS flow as necessary and go to E-1, Loss of Reactor or Secondary Coolant C. Re-initiate Safety Injection from SB HS-27 & 28, SI Actuation Switches, and go to E-0, Reactor Trip or Safety Injection D. Continue in ES-1.1 and align CCP suction to the VCT. Go to E-1, Loss of Reactor or Secondary Coolant if PZR / Charging parameters cannot be maintained Answer: B Explanation:

Adverse containment is present as the MSIVs have just closed due to containment pressure (MSLIS is at 17 psig). Foldout page action for SI reinitiation criteria due to subcooling exist.

A. Incorrect - plausible as restarting one CCP will assist in reestablishing ECCS flow and if it is believed that the containment pressure rise to greater than 17 psig is due to a faulted SG then a foldout page transition to E-2 would be chosen. One CCP was secured in step #2 of ES-1.1 and

NRC Written Examination Callaway Plant Reactor Operator would be available to be started. The choice is incorrect as a faulted SG would cause a rapid rise in CTMT pressure and subcooling to go up due to the rapid cooldown in the affected loop (assuming RCS pressure relatively constant)

B. Correct - As subcooling is below the adverse containment value of 50F, ES-1.1 foldout page states that " IF either condition listed below occurs after SI termination, THEN ESTABLISH ECCS flow as necessary and Go To E-1, Loss Of Reactor Or Secondary Coolant, Step 1:"

C. Incorrect - plausible as this would reestablish ECCS flow but is not the prescribed method and there is no direction to return to E-0 (returning to E-1 is required) making this an incorrect choice.

D. Incorrect - plausible if the adverse containment parameter is not applied to the subcooling value and the decision to continue on in ES-1.1 is made. Realigning CCP suction is step #15 and is the next step that would be performed in the procedure if a foldout page transition is not required Technical Reference(s):

1. ES-1.1, SI Termination, Rev 17 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP#9, Objectives H & I:

H. STATE and EXPLAIN the parameters which are evaluated, including their Criteria and Basis, to transition from ES-1.1 to other procedures.

I. OUTLINE procedural flowpath including major system and equipment operation in accomplishing the goal of ES-1.1.

Question Source: Bank # ___X _ R14978 __

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3____

10 CFR Part 55 Content:

Comments:

Modified the answer by removing boron injection header flow and changed it to ECCS flow to match the wording of ES-1.1 foldout page.

Modified 1 distractor to remove focus on subcooling and added next step in ES-1.1.

The distractor of "Restart the Safety Injection Pumps and go to ES-1.2, Post LOCA Cooldown and Depressurization." was replaced for several reasons.

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Steam Generator Overpressure Group # 2 K/A # W/E13 G2.2.44 Importance Rating 4.2 Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Question #25 The crew entered FR-H.2, Steam Generator Overpressure, due to an overpressure condition on 'B' S/G.

Operators have just completed Step #2 in which Main Feedwater Isolation to the 'B' S/G has been verified

'B' MSIV is closed

'B' S/G narrow range level is 86% steady

'A', 'C', and 'D' S/G NR levels are 5% and slowly rising

AFW flow to each S/G is as follows:

o 'A' S/G 90,000 lbm/hr o 'B' S/G 0 lbm/hr o 'C' S/G 100,000 lbm/hr o 'D' S/G 90,000 lbm/hr Based on the above conditions and in accordance with FR-H.2, what is the NEXT action the crew should take?

A. Transition to FR-H.1, Response to Loss of Secondary Heat Sink B. Locally dump steam from 'B' S/G by establishing SG blowdown C. Dump steam from 'B' S/G using SG ASD or Loop Steam To AFP Turbine D. Locally dump steam from 'B' S/G by opening the Steamline Low Point Drain Valve Answer: C Explanation:

A. Incorrect - this is a red path FR procedure but with the given plant parameters of SG NR level and AFW flow, no red path exists as AFW flow is high enough (>270,000lbm/hr) but plausible as 3 SG NR levels are below 7%.

B. Incorrect - This would be done is additional method to dump steam in step #8 of FR-H.2. This option is not available in step #4 as a method to dump steam from the affected SG but is an

NRC Written Examination Callaway Plant Reactor Operator option to remove pressure later in the procedure and in physical layout of the plant which therefore makes it a plausible choice but not the NEXT action / choice.

C. Correct - step #4 of FR-H.2. Step 4 states "TRY To Dump Steam From Affected SG(s):

SG ASD(s):

o AB PIC-1A (SG A) o AB PIC-2A (SG B) o AB PIC-3A (SG C) o AB PIC-4A (SG D)

OR

MSIV Bypass Valves and condenser steam dump OR

Loop Steam To AFP Turbine:

o AB HIS-5A (SG B) o AB HIS-6A (SG C)

The MSIV Bypass Valves and condenser steam dump option was not included as the stem states that B MSIV is closed and RCS Temp is 550F which is P-12 and would disarm the steam dumps therefore this option is not available.

D. Incorrect - This would be done is additional method to dump steam in step #8 of FR-H.2. This option is not available in step #4 as a method to dump steam from the affected SG but is an option to remove pressure in step #8 making it plausible but is incorrect as it is not the NEXT option and as 'B' MSIV is closed, the option of Steam Line Drains is not appropriate based on current plant conditions and physical plant design.

Technical Reference(s):

1. FR-H.2, Response to Steam Generator Overpressure, Rev 6 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP #26, Objective I: OUTLINE procedural flowpath including major system and equipment operation in accomplishing the goal of:

1. FR-H.1, Response To Loss Of Secondary Heat Sink.

2. FR-H.2, Response To Steam Generator Overpressure Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____3_____

10 CFR Part 55 Content:

NRC Written Examination Callaway Plant Reactor Operator Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 Containment Flooding Group # 2 K/A # W/E15 EA1.3 Importance Rating 2.8 Ability to operate and / or monitor the following as they apply to the (Containment Flooding):

Desired operating results during abnormal and emergency situations.

Question #26 A LOCA has occurred

An ORANGE Path has developed on Containment Critical Safety Function due to Sump level

All Auto Actions have occurred

Annunciator 51D, CCW Surge Tank A Lev High/Low, is LIT

Containment Pressure peaked at 25 psig In accordance with FR-Z.2, Response to Containment Flooding, what would cause this condition?

A. Service Water Leak inside Containment B. Fire Protection System Leak inside Containment C. Component Cooling Water Leak inside Containment D. Containment Spray Line Rupture inside Containment Answer: C Explanation:

A. Incorrect. ESW is the supply for Ctmt loads. Service water is isolated.

B. Incorrect. FP Does supply components in Ctmt. FP alarms are not expected for a LOCA.

C. Correct. FR-Z.2 step 1.c D. Incorrect. Containment Spray actuates at 27 psig. Setpoint not reached Ctmt Sump level of 106" = Orange Path Technical Reference(s):

1. FR-Z.2, Response to Containment Flooding, Rev 007 References to be provided to applicants during examination: None

NRC Written Examination Callaway Plant Reactor Operator Learning Objective: T61.003D, Emergency Operations, LP#30, Objective N: OUTLINE procedural flowpath including major system and equipment operation in accomplishing the goal of FR-Z.1, FR-Z.2, and FR-Z.3 Question Source: Bank # __X L15787___

Modified Bank # ______

New _______

Question History: Last NRC Exam ____2009________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3______

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 1 LOCA CooldownDepressurization Group # 2 K/A # W/E 03 EK2.2 Importance Rating 3.7 Knowledge of the interrelations between the (LOCA Cooldown and Depressurization) 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 #27 The plant has experienced a Small Break LOCA and the crew is performing ES-1.2, Post LOCA Cooldown And Depressurization.

All Reactor Coolant Pumps (RCP) are running

RCS pressure is 1620 psig

An RCS cooldown has been initiated by dumping steam to the atmosphere What describes the optimum RCP configuration, and the basis for this configuration?

A. All RCPs should be stopped; minimize RCS inventory loss when the break uncovers B. Secure all RCPs not required for normal spray flow; minimizes RCS heat input, and still produces effective heat transfer and RCS pressure control C. Secure all RCPs not required for normal spray flow; produces effective heat transfer, provides boron mixing for RHR operations and provides RCS pressure control D. All RCPs should be left running; ensures symmetric heat transfer to the S/Gs, aids in RCS pressure control and prevents steam voiding in the reactor vessel head Answer: B Explanation:

Per the basis for ES-1.2 Step #13, "Forced coolant flow is the preferred mode of operation to allow for normal RCS cooldown and provide PZR spray. If RCPs had not been tripped, all RCPs not required for normal PZR spray are now stopped to minimize heat input to the RCS."

A. Incorrect - plausible as this is the basis for evaluating and securing ECCS pumps during subsequent steps ex Step #14. The basis for securing a CCP states "some ECCS flow is

NRC Written Examination Callaway Plant Reactor Operator necessary to maintain coolant inventory and pressurize the RCS sufficiently to promote primary-to-secondary heat transfer. A conflict arises between keeping the ECCS pumps running to maintain adequate coolant inventory and reducing ECCS flow to minimize leakage from the RCS."

B. Correct - per the basis information provided above C. Incorrect - plausible if the operator believes that RCP D, B & A should be left in service for RCS pressure control as these RCPs provide for normal PZR spray differential pressure used for pressure control in the A and B PZR spray subsystems. A Note prior to step #13 in ES-1.2 states

" NOTE - RCPS should be run in order of priority to provide normal PZR: RCP D, RCP A or RCP B, RCP C." which may be incorrectly applied.

D. Incorrect - plausible if it is believed the priority is maintaining a path to / use of the secondary side heat sink for decay heat removal. Additionally if it is believed that upper head voiding during RCS depressurization must be avoided by leaving all RCPs in service, this is a plausible choice.

The Concern about upper heading voiding is in a NOTE prior to ES-1.2 step #12 which states "NOTE - The upper head region may void during RCS depressurization if RCPs are NOT running.

This will result in a rapidly rising PZR level."

Technical Reference(s):

1. ES-1.2, Post LOCA Cooldown and Depressurization, Rev 19

2. BD-ES-1.2, Basis Document for ES-1.2 Post LOCA Cooldown and Depressurization Rev 12 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP#10, Objective I: OUTLINE procedural flowpath including major system and equipment operation in accomplishing the goal of ES-1.2, Post LOCA Cooldown and Depressurization.

Question Source: Bank # __X L16248____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Reactor Coolant Pump Group # 1 K/A # 003 K1.03 Importance Rating 3.3 Knowledge of the physical connections and/or cause-effect relationships between the RCPS and the following systems: RCP seal system REFERENCE PROVIDED Question #28 All reactor coolant pumps are secured with the following plant conditions:

RCS WR Pressure (BB PI-405) 400 psig

RCS WR Pressure (BB PI-406) 350 psig

Charging Header Pressure (BG PI-120A) 725 psig

VCT Pressure (BG PI-115) 25 psig What is the MAXIMUM #1 seal leak-off flow rate that would allow a reactor coolant pump to be started?

A. 1.0 gpm B. 1.5 gpm C. 2.0 gpm D. 2.5 gpm Answer: D Explanation:

Per attachment 3 of OTN-BB-00003, seal DP is 700 psig which means 2.5 gpm is the highest of the normal operating range.

The distractors are plausible as follows:

If the highest WR Pressure (BB PI-405) was used for the dp and the VCT pressure was subtracted, 400 psig - 25 psig = 375 psig and per the attachment 1.5 gpm would be the maximum allowed.

If the lowest WR Pressure (BB PI-406) was used for the dp and the VCT pressure was subtracted, 350 psig - 25 psig = 325 psig and per the attachment 1.0 gpm would be the maximum allowed.

If the graph is misread and/or the applicant believes than the graph at 700 is not half way between 2 and 3 then 2.0 gpm would be chosen. This value also provides balance as all 4 choices are 0.5 gpm different and in order.

NRC Written Examination Callaway Plant Reactor Operator A. Incorrect - See above explanation B. Incorrect - See above explanation C. Incorrect - See above explanation D. Correct Technical Reference(s):

1. OTN-BB-00003, Reactor Coolant Pumps, Rev 28 References to be provided to applicants during examination:

1. OTN-BB-00003, Reactor Coolant Pumps, Rev 28 Attachment 3 modified by removing the equations at the bottom of the figure.

Learning Objective: T61.0010 Systems, LP #9, Objective Y: EXPLAIN the precautions, limitations and bases for the following components/conditions associated with OTN-BB-00003, "Reactor Coolant Pumps":

1. CCW flow to Thermal Barriers with RCS temperature > 160°F

2. Seal Leakoff Valves with RCS pressure < 100 psig

3. Seal Injection flow

4. RCP starting conditions

5. #1 seal cooling

6. Mode 4 RCP operability

7. RCP starting limitations Question Source: Bank # ______

Modified Bank # __X L6294____

New _______

Question History: Last NRC Exam ____2009________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____2____

10 CFR Part 55 Content:

Comments:

Modified the charging header pressure and VCT pressure changing (from 575 and 50 psig respectively) to the above values therefore the correct answer was changed to 2.5 gpm from 2.0 gpm.

OTN-BB-00003 Rev. 028 Attachment 3 Reactor Coolant Pump Seal Figure Sheet 1 of 1 Page 30 of 30 CONTINUOUS USE

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Reactor Coolant Pump Group # 1 K/A # 003 A3.05 Importance Rating 2.7 Ability to monitor automatic operation of the RCPS, including: RCP lube oil and bearing lift pumps Question #29 When starting 'A' RCP per OTN-BB-00003, Reactor Coolant Pumps, a minimum oil lift system pressure of _____(1)____ is required.

And The interlock is verified satisfied by observing ____(2)____ light on BB HIS-41, RCP A LIFT PUMP, is lit.

A. (1) 200 psi (2) Red B. (1) 200 psi (2) White C. (1) 600 psi (2) Red D. (1) 600 psi (2) White Answer: D Explanation:

Per OTN-BB-00003 note in step "NOTE The Reactor Coolant Pump lift oil interlock is satisfied when the white light on BB HIS-41, RCP A LIFT PUMP, is lit."

Additionally, per Attachment 1 of OTN-BB-00003, a minimum oil lift system pressure of 600 psi is required.

200 psi is plausible because if seal leakoff flow is low the oil lift system is used in the process of addressing the issue per OTN-BB-00003 step 5.1.9 and could be remembered as the interlock setpoint. Red light Lit is plausible as the this is the normal color for running equipment or open valves representing that energy is present.

A. Incorrect - See above explanation - both are wrong

NRC Written Examination Callaway Plant Reactor Operator B. Incorrect - See above explanation - wrong minimum oil lift pressure C. Incorrect - See above explanation - wrong light indication D. Correct Technical Reference(s):

1. OTN-BB-00003, Reactor Coolant Pumps, Rev 28 References to be provided to applicants during examination: None Learning Objective: T61.0010 Systems, LP #9, Objective F: DESCRIBE the purpose and operation of the following RCP components:

12. Oil Lift Pump Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____4____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 CVCS Group # 1 K/A # 004 K1.19 Importance Rating 2.7 Knowledge of the physical connections and/or cause-effect relationships between the CVCS and the following systems: Primary grade water supply Question #30 The Reactor Operator is making a blended flow addition to the RCS and sets up the controls as follows:

Makeup Mode Selector Switch, BG HS-25, to MANUAL

BG FK-110 is set at desired flowrate and in AUTO

BG FK-111 is set at desired flowrate and in AUTO

Total Flow Counter set for 400 gallons

Boric Acid Flow Counter set for 70 gallons

BG HIS-111B (VCT Inlet) is in hard CLOSE

BG HIS-110B (VCT Outlet) is in OPEN Then, the Makeup Water Control Switch, BG HS-26, is placed in RUN.

Due to a malfunction, the Total Flow Counter stops counting the gallons added after 300 gallons of total flow addition.

Assuming NO ACTION by the operator, what is the effect of this malfunction?

A. Boric Acid will continue to inject causing a boration of the RCS B. Reactor Makeup Water will continue to inject causing a dilution of the RCS C. Boric Acid and Reactor Makeup Water will both stop injecting into the RCS immediately D. Boric Acid and Reactor Makeup Water will continue to inject until operator action is taken Answer: B Explanation:

With the Total Flow Counter not counting past 300 gallons, the RMCS will keep adding BA and Reactor Makeup Water until BA counter reached its value, then BA will stop and only Reactor Makeup Water will be added causing a dilution event with no operator action taken

NRC Written Examination Callaway Plant Reactor Operator A. Incorrect see above explanation. Plausible if it is believed that the malfunction will cause Reactor Makeup Water to stop not Boric Acid.

B. Correct C. Incorrect see above explanation. Plausible if it is believed that the both stop due to the malfunction D. Incorrect see above explanation. Plausible if it is believed the effect is to allow the system to continue to operate as the counter would never reach 400 (it stopped at 300) and manual action is required to stop the blended flow addition Technical Reference(s):

1. OTN-BG-00002, Reactor Makeup Control And Boron Thermal Regeneration System, Rev 051 References to be provided to applicants during examination: None Learning Objective: T61.0110, systems, LP #11, Objective F: DESCRIBE the purpose and operation of the following Reactor Makeup Water (RMW) System components:

1. Reactor Makeup Water Storage Tank

2. Reactor Makeup Water Transfer Pumps

3. Chemical Addition Tank

4. Boric Acid Tanks

5. Boric Acid Transfer Pumps

6. Batching Tank

7. BGFCV0111A

8. BGFCV0111B

9. BGFCV0110A

10. BGFCV0110B

11. Blending Tee Question Source: Bank # _X___L5800___

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Residual Heat Removal Group # 1 K/A # 005 K3.07 Importance Rating 3.2 Knowledge of the effect that a loss or malfunction of the RHRS will have on the following:

Refueling operations Question #31 The plant is in Mode 6 with in vessel Fuel movement in progress.

Water level is being maintain at >23 feet above the flange

NB02 bus is de-energized for maintenance activities

'A' RHR pump has tripped and cannot be re-started Per OTO-EJ-00001, Loss of RHR flow, Attachment B, Loss of RHR with Level Greater than 94", what is the FIRST action the crew should take?

A. Start all CTMT Coolers with SW/ESW flow in Slow Speed B. Ensure two SGs are available with a WR level greater than 86%

C. Use Condenser Steam Dumps to maintain RCS temperature stable D. Align Firewater to all available SGs per EOP Addendum 32, Establishing Emergency Feedwater from Fire Water Answer: A Explanation: OTO-EJ-00001, Loss of RHR Flow step 7 sends crew to Attachment B. Att. B step 1 RNO sends crew to step B8 since in mode 6. Step B9 RNO sends crew to step B15 since B train of RHR not available. Step 15.b states: 'Start all CTMT Coolers with SW/ESW flow in Slow Speed" A. Correct Step B15.b B. Incorrect plausible if student does not apply RNO in step B1 and proceeds to Step B3 and applies the RNO which states "ENSURE two SGs are available with: GREATER THAN 86% WR Level". The RCS Loops filled part of the RNO was left out to aid in this distractor plausibility as the RCS loops would be filled for refueling.

C. Incorrect plausible if student does not apply RNO in step B1 and proceeds to Step B4 which includes several actions such as "EOP Addendum 32, Establishing Emergency Feedwater From Fire Water" and Step B4.c "Use Condenser Steam Dumps to control RCS temperature." These are both Mode 5 actions but not procedurally applicable in the situation presented. "Use

NRC Written Examination Callaway Plant Reactor Operator Condenser Steam Dumps to control RCS temperature is also step B10 which would be reached if the applicant does not correctly apply the RNO at step B9 branching to step B15.

D. Incorrect -See explanation for Distractor C Technical Reference(s):

1. OTO-EJ-00001, Loss of RHR Flow, Rev 038 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP#62, Objective E:Given a set of conditions, DISCUSS the required flowpath to stabilize the plant for a loss of RHR per OTO-EJ-00001, Loss of RHR Flow.

Question Source: Bank # ______

Modified Bank # ____

New ___X____

Question History: Last NRC Exam _____NA_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

LOD ____3______

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Residual Heat Removal Group # 1 K/A # 005 A4.02 Importance Rating 3.4 Ability to manually operate and/or monitor in the control room: Heat exchanger bypass flow control Question #32 Residual Heat Removal Train B is in service in a cooldown lineup with the following conditions:

EJ FK-619, RHR HX B BYPASS CTRL, is in AUTO controlling flow at 3500 gpm as indicated on EJ FI-619, RHR TO ACC INJ LOOPS 3 & 4 FLOW

EJ HCV-607, RHR HX B FLOW CTRL VLV is 15% open Flow transmitter EJ FT-619 fails LOW. How do the following valves respond?

EJ FCV-619, RHR HX B Bypass Flow Control Valve, ____(1)_____ and EJ FCV-611, RHR PUMP B Miniflow Valve, is ____(2)____.

A. (1) Opens (2) Open B. (1) Opens (2) Closed C. (1) Closes (2) Open D. (1) Closes (2) Closed Answer: B Explanation:

EJ FCV-619 opens on lowering flow sensed at EJ FT-619 to raise flow. EJ FCV-611 operates off flow transmitter EJ FT-611 at the discharge of the pump and prior to the HX. With flow already at 3500 gpm, this valve is closed. The valve will remain closed unless flow lowers to 816 gpm as sensed at EJ FT-611. Since flow only rises due to the EJ FT-619 failure and EJ FCV-619 response, EJ FT-611 only sees more flow than prior to failure. Thus, EJ FCV-611 remains closed.

NRC Written Examination Callaway Plant Reactor Operator A. Incorrect - part 2 is wrong - EJ FCV-611 position is wrong but plausible if examinee thinks EJ FCV-611 utilizes EJ-FT-619 for control.

B. Correct - See above explanation C. Incorrect - both are wrong. EJ FCV-619 closing is plausible if examinee doesn't understand reverse acting controller and believes it should close. As stated above, EJ FCV-611 position is wrong but plausible if examinee thinks EJ FCV-611 utilizes EJ-FT-619 for control.

D. Incorrect - part 1 is incorrect. As stated above, EJ FCV-619 closing is plausible if examinee doesn't understand reverse acting controller and believes it should close Technical Reference(s):

1. M-22EJ01, P&ID RHR System, Rev 62 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#7, Objective B: DESCRIBE the purpose and operation of the following RHR System components, to include interlocks, controller operation and power supplies:

6. RHR Heat Exchanger Flow Control Valves

7. RHR Heat Exchanger Bypass Valves Question Source: Bank # __X L23368____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____2______

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Emergency Core Cooling Group # 1 K/A # 006 K4.11 Importance Rating 3.9 Knowledge of ECCS design feature(s) and/or interlock(s) which provide for the following: Reset of SIS Question #33 To completely reset a valid Safety Injection Actuation following a LOCA, what are the time and Reactor Trip Breaker (RTB) position requirements?

A. Any time after actuation One RTB open ONLY B. One minute after actuation One RTB open ONLY C. Any time after actuation BOTH RTBs open D. One minute after actuation BOTH RTBs open Answer: D Explanation: Per Dwg 7250D64 S008 there is a one minute time delay to reset a SIAS. Also there is a reset seal in circuit from the RTBs. If RTB is not open then that train will not reset if SI signal still active. Both RTBs have to be open when resetting SIAS to get a complete reset A. Incorrect - see above explanation B. Correct - see above explanation C. Incorrect - see above explanation D. Incorrect - see above explanation Technical Reference(s):

1. 7250D64 S002, Reactor Trip Signals - Functional Diagram, Rev 006

2. 7250D64 S008, Safeguards Actuation Signals - Functional Diagram, Rev 004 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#17, Objective D:STATE the conditions that will

NRC Written Examination Callaway Plant Reactor Operator initiate a Safety Injection Signal and DESCRIBE the conditions necessary to reset the signal.

Question Source: Bank # ______

Modified Bank # ______

New __X_____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD _____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Pressurizer Relief/Quench Tank Group # 1 K/A # 007 G2.1.23 Importance Rating 4.3 Ability to perform specific system and integrated plant procedures during all modes of plant operation Question #34 Reactor Power is 100% when the following alarm is received:

34E, PRT PRESS HI PRT parameters are as follows:

Pressure is 9 psig and RISING SLOWLY.

Level is 67% and STABLE.

If allowed to continue, what is the potential impact of this event and the action required to restore PRT pressure?

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

And (2) Per OTN-BB-00004, Pressurizer Relief Tank, _____(2)______ to prevent PRT rupture disc operation.

A. (1) 50 psig (2) vent the PRT B. (1) 50 psig (2) drain the PRT to the containment sump C. (1) 100 psig (2) vent the PRT D. (1) 100 psig (2) drain the PRT to the containment sump Answer: C

NRC Written Examination Callaway Plant Reactor Operator Explanation: Per OTN-BB-00004, rupture disc breaks at 100 psig and draining PRT is for level control not pressure. Venting PRT is for lowering pressure..50 psig is the maximum pressure following a design discharge into the PRT A. Incorrect see above explanation B. Incorrect see above explanation C. Correct see above explanation D. Incorrect see above explanation Technical Reference(s):

1. OTN-BB-00004, Pressurizer Relief Tank, Rev 041 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#9, Objective U: EXPLAIN the precautions, limitations and bases for the following processes/conditions associated with OTN-BB-00004, "Pressurizer Relief Tank":

PRT TEMP HI alarm

Normal tank level

Nitrogen blanket

PRT venting for refueling Question Source: Bank # __X__L16178___

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2005_____

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis _X __

LOD _____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Component Cooling Water Group # 1 K/A # 008 A2.01 Importance Rating 3.3 Ability to (a) predict the impacts of the following malfunctions or operations on the CCWS, and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of CCW pump Question #35 Reactor Power is 100%.

'A' CCW Pump is in service

'B' and 'D' CCW Pumps are in AUTO

'C' CCW Pump is Pull To Lock and unavailable

'A' Train CCW is aligned to the Service Loop Then, the following alarms are received:

52B, CCW Pump A/C Press LO

51B, CCW Pump A/C Trouble

51C, CCW PMP A Flow LO

51D, CCW SRG TK A Lev HILO

'A' CCW Surge Tank Level is 35%.

The crew has just entered OTO-EG-00001, CCW System Malfunction.

Based on the above conditions, what is the FIRST action(s) the crew should perform per OTO-EG-00001?

A. Verify the Reactor is tripped and trip the RCPs B. Align ESW Makeup to the 'A' CCW Surge Tank C. Transfer the service loop from Train A to Train B D. Ensure DI water to CCW Surge Tank A, EGLV0001, is Closed Answer: C

NRC Written Examination Callaway Plant Reactor Operator Explanation: Note: there are no immediate actions associated with OTO-EG-00001, just continuous action steps based on time.

Per OTO-EG-00001, Steps #8 and #9 IF CCW surge tank is steady or rising, the #8 RNO applies and since A tank is below the point in which auto makeup occurs (43.75% per the annunciator) it is expected that the A makeup valve should be open (no closed as in the distractor), but plausible as the candidate may not correctly remember the auto makeup setpoint and see that level is steady, and therefore believe the next action is to verify this valve is closed.

Specifically step RNO #8 b "IF CCW Surge Tank A is rising AND makeup is NOT required, THEN PERFORM the following: 1) ENSURE EGLV0001, DI CCW Surge Tank A is closed:

The distractor of align ESW makeup to the 'A' surge tanks is plausible per OTO Step #9 RNO as the tank level is below 44% (auto makeup setpoint in the OTO or 43.75% in the annunciator) and not rising.

The action to direct to manually trip the reactor and RCP is from continuous step #10 if the training supplying the service loop surge tank level is less than 10%. Additionally per continuous action step #5 if CCW is lost to the RCPs for greater than 10 minutes, operators are required to trip the reactor and trip the RCPs making it a plausible distractor. However, this is not the FIRST action the crew would perform. CCW service train realignment per step #1 RNO should be completed within 10 minutes. Tripping the reactor and RCPs is not the FIRST action in the procedure for the given stem conditions. In fact step #1 RNO says do not proceed until the service loop has been transferred.

With the pump configuration given and the annunciators provided, the A Train of CCW will not have a running CCW pump (due to the loss of the A pump and the C pump in PTL) and per the RNO of step #1 b 1) TRANSFER the Service Loop to alternate Train using one following:

Attachment C, Transferring Service Loop From Train A to Train B. The annunciators provided show that the 'A' CCW pump tripped and with the 'C' pump in PTL, the operator is required to predict the impact of the A CCW pump trip i.e. no 'A' Train CCW pumps will be running.

A. Incorrect - See explanation above B. Incorrect - See explanation above C. Correct - See explanation above. The First action the crew should take per the stem conditions is to transfer the service loop per step #1 RNO as the goal of the procedure is to restore CCW to the service loop prior to the need secure RCPs per the timed continuous action step #4 and #5.

D. Incorrect - See explanation above Technical Reference(s):

1. OTO-EG-00001, CCW System Malfunction, Rev 18

2. OTA-RK-00020, ADD 52B, CCW Pump A or C Pressure Low, Rev 0

3. OTA-RK-00020, ADD 51B, CCW Pump A/C Trouble, Rev 3

4. OTA-RK-00020, ADD 51D, CCW Surge Tank A Level High/Low, Rev 1 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP #17, Objective E: Given a set of plant conditions or parameters indicating a CCW System Malfunction, ANALYZE the correct

NRC Written Examination Callaway Plant Reactor Operator procedure(s) to be utilized and the required actions to stabilize the plant.

Question Source: Bank # _X__(No ID)____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____NA________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____2_____

10 CFR Part 55 Content:

Comments:

The Question was written to part B of the K/A as the ability to predict the impacts the loss of a CCW pump is inherent to part b in which the operator must determine the correct course of action to address the event.

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Component Cooling Water Group # 1 K/A # 008 A1.02 Importance Rating 2.9 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CCWS controls including: CCW temperature Question #36 During a plant cooldown on RHR, what is the MAXIMUM CCW heat exchanger outlet temperature ___(1)____ and time ___(2)___ allowed.

A. (1) 105°F (2) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months B. (1) 120°F (2) 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months C. (1) 105°F (2) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D. (1) 120°F (2) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Answer D Explanation: : Per OTN-EG-00001, during normal plant operation, the CCW Heat Exchanger outlet temperature should not be allowed to exceed 105°F (120°F is allowed during approximately the first four hours of RHR operation for cooldown).

A. Incorrect see above explanation B. Correct see above explanation C. Incorrect see above explanation D. Incorrect see above explanation Technical Reference(s):

1. OTN-EG-00001, Component Cooling Water system, Rev 062 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#10, Objective G: DISCUSS significant precautions

NRC Written Examination Callaway Plant Reactor Operator and limitations associated with operating the CCW System per OTN-EG-00001, "Component Cooling Water System Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3______

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Pressurizer Pressure Control Group # 1 K/A # 010 G2.1.32 Importance Rating 3.8 Ability to explain and apply system limits and precautions Question #37 The plant is in Mode 3, at normal operating pressure and temperature.

The crew is initiating auxiliary spray in accordance with OTN-BB-00005, Pressurizer and Pressurizer Pressure Control. The operator is verifying spray line deltaT is acceptable.

What is the MINIMUM regenerative heat exchanger outlet temperature allowable for Auxiliary Spray initiation under these conditions and why?

A. 300°F To prevent excessive spray line stress B. 340°F To prevent excessive spray line stress C. 300°F To minimize pressurizer pressure transient D. 340°F To minimize pressurizer pressure transient Answer: B Explanation: Per OTN-BB-00005, the PZR vapor space to Regen Hx outlet delta T needs to be less than 320°F. At NOP, PZR temperature is 653°F minus a max T of 320°F equals a minimum of 333°F for regenerative heat exchanger outlet temperature. With a regenerative heat exchanger outlet temperature of 340°F, the T is 313°F and with the regenerative heat exchanger outlet temperature of 300°F, the T is 353°F, which is greater than the allowable T of 320°T. This prevents excessive spray line stress per FSAR 16.4.4.1 A. Incorrect see above explanation B. Correct see above explanation C. Incorrect see above explanation D. Incorrect see above explanation

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

1. OTN-BB-00005, Pressurizer and Pressurizer Pressure Control, Rev 16

2. FSAR 16.4.4.1 References to be provided to applicants during examination:

1. ASME Steam Tables Compact Edition, Vo; 83 Learning Objective: T61.0110 Systems, LP#9, Objective W: EXPLAIN the precautions, limitations and bases for the following:

1. OTN-BB-00005, "Pressurizer and Pressurizer Pressure Control" Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ___N/A_____

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

LOD _____4_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Reactor Protection Group # 1 K/A # 012 K4.02 Importance Rating 3.9 Knowledge of RPS design feature(s) and/or interlock(s) which provide for the following:

Automatic reactor trip when RPS setpoints are exceeded for each RPS function; basis for each Question #38 The Reactor Protective System (RPS) setpoint for Overpower T has been exceeded.

(1) The Reactor Trip Breaker's shunt trip coils ___(1)___ to open the breaker?

And (2) The Overpower T trip function (2).

A. (1) energize (2) ensures that the allowable heat generation rate (kw/ft) of the fuel is not exceeded B. (1) energize (2) ensures that the design limit Departure from Nucleate Boiling Ratio (DNBR) is not exceeded C. (1) deenergize (2) ensures that the allowable heat generation rate (kw/ft) of the fuel is not exceeded D. (1) deenergize (2) ensures that the design limit Departure from Nucleate Boiling Ratio (DNBR) is not exceeded Answer: A Explanation:

For reactor trip actuations, the logic bay causes the associated trip breaker's UV coil to de-energize (and the shunt coil to energize) which opens the trip breaker, dropping the control rods.

Note the shunt coils receive power from NK to energize the shunt trip device.

NRC Written Examination Callaway Plant Reactor Operator Therefore 'deenergize' is a plausible distractor as the undervoltage coils deenergize upon a RPS signal.

The Overpower delta T trip Function ensures that the allowable heat generation rate (kW/ft) of the fuel is not exceeded. The overtemperature delta T trip Function is provided to ensure that the design limit DNBR is met and is a plausible distractor as it is the basis for a separate RPS function. Per the FSAR page 7.2-6 "Overpower DT trip This trip protects against excessive power (fuel rod rating protection) and trips the reactor on coincidence" A. Correct B. Incorrect - the reason is wrong C. Incorrect - both are wrong D. Incorrect - deenergized is wrong Technical Reference(s):

1. E-23SB10A, RT Switchgear Train A, Rev 1

2. Callaway FSAR, Section 7.2.1.1.2 Reactor Trips" References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems, LP #27, Objective A: STATE the function and EXPLAIN the design criteria of the Reactor Protection System (RPS).

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Reactor Protection Group # 1 K/A # 012 A2.07 Importance Rating 3.2 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: Loss of dc control power Question #39 Reactor Power is 100%.

OSP-SB-0001B, Reactor Trip Breaker (RTB) 'B' Trip Actuating Device Operational Test is in progress

During the test, the Undervoltage (UV) coil on RTB 'B' failed to actuate

RTB 'A' and Reactor Trip Bypass Breaker (BYP) 'B' are RACKED IN and CLOSED A common cause analysis determined that a parts issue makes the UV coils suspect on all reactor trip and bypass breakers.

Then the following occurs:

DC Bus NK01 is lost

A manual Reactor Trip is initiated from SB HS-1 and SB HS-42

E-0, Reactor Trip or Safety Injection, is entered

The UV coils on BOTH RTB 'A' & BYP 'B' fail to actuate Based on the above conditions, the crew will A. remain in E-0. ONLY BYP 'B' opened on a shunt trip B. remain in E-0. ONLY RTB 'A' opened on a shunt trip C. remain in E-0. BOTH RTB 'A' and BYP 'B' opened on a shunt trip D. immediately transition to FR-S.1, Response to Nuclear Power Generation/ATWS.

NEITHER RTB 'A' nor BYP 'B' is open Answer: A Explanation:

NRC Written Examination Callaway Plant Reactor Operator A. Correct, RTBB 'B' shunt trip coil is operated by DC bus NK02. Since DC bus NK02 is still energized. RTBB 'B' will open. Crew should stay in E-0.

B. Incorrect, the shunt trip coil for RTB 'A' requires power from DC bus NK01 to operate.

C. Incorrect, the shunt trip coil for RTB 'A' requires power from DC bus NK01 to operate.

D. Incorrect, RTB 'A' will remain closed, but RTBB 'B' will open and E-0 has actions for opening MG set supply breakers Technical Reference(s):

1. E-0, Reactor Trip or Safety Injection, Rev 026

2. E-21NK02, Class IE 125v DC System - Meter & Relay Diagram, Rev 019 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#27, Objective B: DRAW, LABEL and EXPLAIN a simplified diagram of the RPS from the sensor to the Solid State Protection System (SSPS).

Question Source: Bank # ______

Modified Bank # ______

New __X____

Question History: Last NRC Exam _____NA_____

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____4_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Engineered Safety Features Actuation Group # 1 K/A # 013 K2.01 Importance Rating 3.6 Knowledge of bus power supplies to the following: ESFAS/safeguards equipment control Question #40 What is the power supply to SA066B, ESFAS Logic Cabinet?

A. NN01 B. NN02 C. NN03 D. NN04 Answer: D Explanation:

Per OOA-SA-C066X step 4.1 and E23-SA22, the power supply to SA066B is NN0421. All other 120VAC NN buses are plausible as they are all safety related NN buses but do not supply 120VAC power to this EFSAS Logic Cabinet.

OOA-SA-C006X step 4.1 states "The power supplies to the ESF logic/termination racks are:

SA066A - NN0104

SA066B - NN0421

SA066C - NN0219" A. Incorrect B. Incorrect C. Incorrect D. Correct Technical Reference(s):

1. OTS-SA-00001, Operations of EFSAS, Rev 19

2. E-23SA22, Schematic Diagram, ESFAS Cabinets, Rev 1

3. OOA-SA-C066X, EFSAE Panel SA066X Alarm Information, Rev 17 References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems, LP #52, EFSAS, Objective E & C & F:

NRC Written Examination Callaway Plant Reactor Operator F: DISCUSS the purpose and scope of the following: OTS-SA-00001, "De-energizing and Energizing Engineered Safety Feature Actuation System.

E: IDENTIFY the ESFAS status panel controls, alarms and indications and DESCRIBE how each is used to predict, monitor, test or control the ESFAS.

C; DISCUSS the following concerning the ESFAS power up/down sequence:

1. The purpose of blocking crosstrips from the de-energized channel prior to de-energization.

2. De-energizing the 48 VDC output relay power before de-energizing the dual voltage electronics power supply on a down power.

3. Ensuring that all actuations are reset prior to energizing the 48 VDC output relay power on an up power.

Question Source: Bank # __X____

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2016 ILT Exam_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Containment Cooling Group # 1 K/A # 022 K3.01 Importance Rating 2.9 Knowledge of the effect that a loss or malfunction of the CCS will have on the following:

Containment equipment subject to damage by high or low temperature, humidity, and pressure Question #41 To protect the CRDM coils, what is the ambient temperature limit in the CRDM shroud on a loss of CRDM cooling?

A. 120°F B. 140°F C. 165°F D. 175°F Answer:C Explanation:

A. Incorrect plausible since temperature to protect area below pressurizer skirt B. Incorrect plausible since temperature to protect cavity concrete temperature C. Correct temperature limit for CRDM Shroud D. incorrect plausible since it corresponds to RCS when a minimum of, either Containment Cooler Fan A or C AND one Containment Cavity Cooling Fan should be in operation Technical Reference(s):

1. OTN-GN-00001 Containment Cooling and CRDM Cooling, Rev 030

2. OTO-GN-00002, CRDM Cooling Fan Malfunction, Rev 006 References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems, LP#40, Objective O: EXPLAIN the precautions, limitations and bases for the following processes/conditions associated with OTN-GN-00001, "Containment and CRDM Cooling":

1. RCS temperature and Containment Fan Cooler and Cavity Cooling Fan operation

2. RCS temperature and CRDM fan operation

3. Cavity Cooling Fans and cavity concrete temperature

4. Pressurizer Cooling Fan or Containment Cooler 1D operation

NRC Written Examination Callaway Plant Reactor Operator

5. Containment Cooler SW/ESW flow restoration

6. Slow speed operation for Containment Coolers Question Source: Bank # ______

Modified Bank # __X L5589__

New _______

Question History: Last NRC Exam ___NA_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____4_____

10 CFR Part 55 Content:

10 CFR 55.41(b)(9)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Containment Spray Group # 1 K/A # 026 A1.01 Importance Rating 3.9 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the CSS controls including: Containment pressure Question #42 At Time = 0, a Large Break LOCA occurred coincident with a Loss of Off-Site Power.

All systems responded as designed

At T= 12 seconds, both EDG output breakers close

At T= 20 seconds, Containment pressure has reached 9 psig If the Containment pressure trend continues, what is the earliest Time, T=_____

seconds, the CS pumps will auto start?

A. 27 B. 53 C. 60 D. 75 Answer: C Explanation:

The correct answer is correct based on a pressure rise of 9 psig per 20 seconds, the auto actuation setpoint of 27 psig will be reached in 1 minute (T=60 seconds)

At the time of the event (T=0) the LOCA occurs and D/G is up to speed/voltage and breaker closed at T=12.

CS pumps will receive a permissive at T=27 but containment pressure has not reached the setpoint of 27 psig so no actuation will occur.

CS pumps will obtain another start signal at T=53 but again no actuation will occur since containment pressure has not reached 27 psig.

All distracters are plausible based on the examine not taking into account the assumed 12 seconds to get the DG up to speed/voltage and breaker closed, picking the wrong actuation setpoint or not understanding the normal timer sequence. The pump will start after T=60 assuming valve stroke time delay.

A. Incorrect, CSAS actuation not reached B. Incorrect, CSAS actuation not reached

NRC Written Examination Callaway Plant Reactor Operator C. Correct D. Incorrect, CSAS actuation not added to CS Pump start Technical Reference(s):

1. E-22NF01, Load Shedding and Emergency Load Sequencing Logic rev 008 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#51, Objective C DISCUSS the reason for the five (5) second sequencing interval and LIST the major loads sequenced by the following:

1. LOCA Sequencer.

2. Shutdown Sequencer.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam __NA_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Containment Spray System Group # 1 K/A # 026 K3.01 Importance Rating 3.9 Knowledge of the effect that a loss or malfunction of the CSS will have on the following: CSS Question #43 During performance of ECA-1.1, Loss of Emergency Coolant Recirculation, the current plant conditions are:

Annunciator 18A, NB01 Bus Lockout

RWST level is 37%

CTMT Recirc Sump Level is 10 inches

CTMT Pressure is 30 PSIG Per ECA-1.1, what is the MINIMUM required combination for operation of CTMT COOLERS and/or CTMT SPRAY PUMPS running for these conditions?

A. 1 CTMT Cooler and 2 CTMT Spray Pumps B. 2 CTMT Coolers and 1 CTMT Spray Pump C. 2 CTMT Coolers and 0 CTMT Spray Pumps D. 4 CTMT Coolers and 0 CTMT Spray Pumps Answer: B Explanation:

Per ECA-1.1 step 7 (see below) for the given conditions the following combinations would be acceptable, 2 CS pumps and 0/1 coolers, 1 CS pump and 2/3 coolers and 0 CS pumps and 4 coolers. Only 1 CS pump and 2 containment are available

NRC Written Examination Callaway Plant Reactor Operator A. Incorrect plausible if student doesn't recognize only 1 CS pump available B. Correct see table above C. Incorrect plausible if student doesn't recognize only 2 containment coolers available D. Incorrect plausible if student picks wrong pressure on table Technical Reference(s):

1. ECA-1.1 Loss Of Emergency Coolant Recirculation, Rev 014 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP#13, Objective 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 # ______

Modified Bank # __X___L16349_

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3______

10 CFR Part 55 Content:

NRC Written Examination Callaway Plant Reactor Operator Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Main and Reheat System Group # 1 K/A # 039 A1.03 Importance Rating 2.6 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with operating the MRSS controls including: Primary system temperature indications, and required values, during main steam system warm-up Question #44 The plant is in Mode 3 at NOP/NOT with all Shutdown Banks withdrawn.

Steam dumps are in Steam Pressure Mode with a setpoint of 1092 psig.

During the performance of OTN-AB-00001, Main Steam and Steam Dump Systems, section 5.4 System Warmup, an OT inadvertently OPENS all steamline drains downstream of the MSIVs.

(1) If no operator action is taken, the primary system cooldown due to continuing steam flow will stop at a reactor coolant system temperature of approximately And (2) Technical Specifications 3.4.2 (RCS Minimum Temperature for Criticality)

_____________ applicable in Mode 3.

A. (1) 491°F (2) is B. (1) 491°F (2) is NOT C. (1) 550°F (2) is D. (1) 550°F (2) is NOT Answer: B Explanation: P-12 (550°F) will isolate steam flow to the steam dump and would normally stop the cooldown but because of the location of the inadvertently opened steam line drains valves, even when P-12 actuates, the cooldown will continue past that and stop only at the temperature

NRC Written Examination Callaway Plant Reactor Operator value corresponding to a Main Steam Isolation of 615 psig (491°F). The second part is plausible if the applicant believes that withdrawn shutdown banks constitute Mode 2 entry which is normally called at control bank withdrawal.

A Incorrect. First part is correct. The second part is plausible if the applicant believes that withdrawn shutdown banks constitute Mode 2 entry which is normally called at control bank withdrawal.

B. Correct With the conditions given, the cooldown will not be secured until a Main Steam Isolation occurs at 615 psig. This corresponds to 491°F. T.S. 3.4.2 is applicable in Modes 1 and 2 with Keff 1.0 which is not entered until control bank rods are withdrawn.

C. Incorrect. Both are wrong. First part is plausible if the applicant believes that actuation of P-12 (550°F) will isolate the steam flow (steam dumps) and stop the cooldown. The second part is plausible if the applicant believes that withdrawn shutdown banks constitute Mode 2 entry which is normally called at control bank withdrawal.

D. Incorrect. First part is incorrect but is plausible if the applicant believes that actuation of P-12 (550°F) will isolate the steam flow (steam dumps) and stop the cooldown.

Technical Reference(s):

1. OTG-ZZ-00001, Plant Heatup Cold Shutdown to Hot Standby, Rev 093

2. OTN-AB-00001, Main Steam and Steam Dump Systems, Rev 023

3. Technical Specifications 3.4.2, Minimum Temperature for Criticality, Rev 202 References to be provided to applicants during examination: None Learning Objective: T61.003A, Normal Operations, LP#4, Objective E. OUTLINE major actions required After Mode 3 Entry while heating RCS to NOP/ NOT

i. DISCUSS coordination of the following activities including major equipment operation and required RCS parameters (Pressure, Temperature, Heatup rate, and Pressurizer Level)

1. Safety Injection System alignment

2. CCP/ CVCS alignment

3. Safety Injection Accumulator alignment

4. RCP operation

5. Pressurizer pressure and level control

6. Enabling Steamline Pressure SI

7. Startup Main Feedwater Pump operation

8. Steam Trap bypass valve operation Question Source: Bank # ______

Modified Bank # _______

New ___X____

Question History: Last NRC Exam ___NA_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____3_____

10 CFR Part 55 Content:

10 CFR 55.41(b)(4)

Comments:

NRC Written Examination Callaway Plant Reactor Operator

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier 2 Main and Reheat Steam Group 1 K/A 039 K5.01 Importance Rating 2.9 Knowledge of the operational implications of the following concepts as they apply to MRSS:

Definition and causes of steam/water hammer Question #45 A plant heatup is in progress in accordance with step 5.7.4 (Initiate main steam warmup) of OTG-ZZ-00001, Plant Heatup From Cold Shutdown to Hot Standby.

What is the next action required, and the reason for that action, in accordance with OTG-ZZ-00001?

A. Periodically cycle low point steam drains during the heatup to prevent water hammer in the main steam lines B. Isolate main steam drains when RCS temperature is above 300°F to prevent water hammer in the main steam lines C. Maintain main steam drains open when RCS temperature is above 300°F to expedite RCS heatup D. Maintain steam trap bypass valves open during the heatup to expedite RCS heatup Answer: A Explanation:

A: CORRECT: - Step 5.7.5 of OTG-ZZ-00001(During RCS Heatup INITIATE Step 5.7.5.a OR 5.7.5.b to prevent water hammer in the Main Steam lines):

B: INCORRECT - because water buildup may water hammer if valves are closed.

C: INCORRECT -:Note prior to step 5.7.5 says "In order to expedite RCS Heatup, Secondary Plant steam drains and steam trap bypasses may be periodically cycled per Attachment 7, Steam Load Management For RCS temperature Control instead of continuously open as instructed in the Normal Operating Procedures" but incorrect because maintaining valves open will delay the heatup.

D: INCORRECT - Note prior to step 5.7.5 says "In order to expedite RCS Heatup, Secondary Plant steam drains and steam trap bypasses may be periodically cycled per Attachment 7, Steam Load Management For RCS temperature Control instead of continuously open as instructed in the Normal Operating Procedures" but incorrect because maintaining valves open will delay the heatup. Additionally a caution prior to step 5.7.5.b discusses the consequences of opening these drains and is not the desired to leave them open.

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

1. OTG-ZZ-00001, Plant Heatup From Cold Shutdown to Hot Standby, Rev 93 References to be provided to applicants during examination: None Learning Objective: T61.GFES, LP #27, Objective #43: Explain operational implications of water (fluid) hammer.

T61.0110 Systems: LP #31, Objective I: DISCUSS the following operating conditions associated with the Main Turbine and Auxiliary Components:

1. Initial Conditions to Admit Steam

2. Turbine Shell Warming

3. Steam Chest Warming

4. Turbine Roll Question Source: Bank # ____X 16251_____

Modified Bank # _____

New _______

Question History: Last NRC Exam ___2019 NRC Q#43____

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

10 CFR 55.41(b)(14)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Main Feedwater Group # 1 K/A # 059 A3.06 Importance Rating 3.2 Ability to monitor automatic operation of the MFW, including: Feedwater isolation Question #46 Reactor Power is 100%.

A malfunction causes all S/G levels to lower to 5% NR level

The operating crew inserted a manual reactor trip during the transient

Auxiliary Feedwater has restored all S/G levels to 51% NR level (1) At SB069, the Lo Lo S/G Level Bistables are currently .?

And (2) What is the status of the S/G FRV Bypass Valves, AEFCV550/560/570/580?

A. (1) LIT (2) The S/G FRV Bypass Valves will indicate RED B. (1) LIT (2) The S/G FRV Bypass Valves will indicate GREEN C. (1) OFF (2) The S/G FRV Bypass Valves will indicate RED D. (1) OFF (2) The S/G FRV Bypass Valves will indicate GREEN Answer: D Explanation: Per OTO-SA-00001, Attachment K, step K.1 and the note prior to this step, the Bistables should be extinguished. This is because " The S/G Lo Lo Level FWIS Does not have a reset. When the S/G Lo Lo Level bistables on SB069 are clear, the S/G Lo Lo Level FWIS will clear automatically." These Bistables and the associated FWIS reset automatically when SG NR level clears the Lo Lo setpoint of 21%. LIT is plausible if the candidate believes that manual action is required to reset the Lo Lo S/G FWIS as all other FWIS require some type of manual operator action to reset the signal.

Also meet conditions for FWIS on P-4 and 564F. This maintains FWIS even though S/G levels restored keeps FRV Bypass valves closed.

NRC Written Examination Callaway Plant Reactor Operator Per OTO-SA-00001, Attachment AL, the S/G FRV Bypass Valves are closed (indicate GREEN) during a FWIS. RED (remaining open) is plausible if the candidate confuses these valves with valves in the SB blowdown system (which close on SGBIS) or confuses the indication they would observe on a control room panel.

A. Incorrect - both are wrong - see above explanation B. Incorrect - the bistables would be off as SG NR level has returned to program level C. Incorrect - The S/G FRV Bypass valves would be Green as they are closed D. Correct - see above explanation Technical Reference(s):

1. OTA-RK-00026, Addendum 126B, Rev 000

2. OTO-SA-00001, ESFAS Verification and Restoration, Rev 042 Attachment AL, AQ,and K

3. 7250D64 S013, Functional Diagram Feedwater Control and Isolation, Rev 15

4. E-0, Reactor Trip or Safety Injection, Rev 25 Step #7 References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems, LP #23, Objective C: STATE the conditions, including the setpoints and coincidences, that will cause a FWIS and EXPLAIN the system response to the signal.

Question Source: Bank # __X (no ID)____

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2016______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3______

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Auxiliary/Emergency Feedwater Group # 1 K/A # 061 A3.01 Importance Rating 4.2 Ability to monitor automatic operation of the AFW, including: AFW startup and flows Question #47 Reactor Power was 100% when a plant trip occurred.

A lockout of ESF Train 'B' 4160 Volt Bus NB02 occurs

The Turbine Driven Auxiliary Feedwater Pump fails to start What 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:

The A MDAFP can feed B and C S/Gs.

The B MDAFP can feed A and D S/Gs The TDAFP can feed all 4 S/Gs.

With the condition given in the stem, the A MDAFP is the only Aux Feedwater pump in service and with a start signal (shrink from a plant trip from 100%) it will be feeding the B and C S/Gs.

A. Incorrect - plausible as it is an combination of S/Gs that provide distractor balance to the question B. Incorrect - plausible as this would be correct if the B MDAFP was the only pump in service C. Correct D. Incorrect - plausible as it is an combination of S/Gs that provide distractor balance to the question Technical Reference(s):

1. M-22AL01, P&ID Auxiliary Feedwater, Rev 50 References to be provided to applicants during examination: None

NRC Written Examination Callaway Plant Reactor Operator Learning Objective: T61.0110 Systems, LP #25, Objective B. DESCRIBE the purpose, operation and location of the following AFW System components:

1. Motor Driven AFW Pumps

2. Turbine Driven AFW Pumps Question Source: Bank # _X__L16896___

Modified Bank # ______

New _______

Question History: Last NRC Exam __2013______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis _X___

LOD ____2_____

10 CFR Part 55 Content:

(10 CFR 55.41(b)(4)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 AC Electrical Distribution Group # 1 K/A # 062 A4.02 Importance Rating 2.5 Ability to manually operate and/or monitor in the control room: Remote racking in and out of breakers Question #48 The following indications exist on NB HIS-2, NB01 NORMAL SPLY BKR NB0112 SWITCH, hand indicating switch:

RED LIGHT LIT

GREEN LIGHT NOT LIT

AMBER LIGHT NOT LIT ESF Transformer XNB01 is ENERGIZED, NB01 Voltmeter indicates 0 Volts What describes the condition of the breaker?

A. Racked IN and OPEN B. Racked IN and CLOSED C. Racked to TEST and OPEN D. Racked to TEST and CLOSED Answer: D Explanation:

Red light will be lit when breaker is in closed position regardless if racked to test or racked in.

When breaker is racked in and closed there will voltage indicated on voltmeter. Green light will be lit when breaker is in open position regardless if racked to test or racked in and no voltage will be indicated on voltmeter.

A. Incorrect see explanation above B. Incorrect see explanation above C. Incorrect see explanation above D. Correct see explanation above Technical Reference(s):

1. E-23NB12, Class 1E Bus NB01 Feeder Breaker 152NB0112, rev 13

NRC Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems, LP#6, Objective C IDENTIFY the Safeguards Power System Main Control Board (MCB) controls and indications and DESCRIBE how each is used to predict, monitor or control changes in the Safeguards Power System.

Question Source: Bank # X____L12009_

Modified Bank # ______

New _______

Question History: Last NRC Exam ___NA_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

LOD _____2_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 DC Electrical Distribution Group # 1 K/A # 063 G2.4.26 Importance Rating 4.2 Ability to verify that the alarms are consistent with the plant conditions.

Question #49 Reactor power is 100% when the following indications are observed by the crew:

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

____(1)____, and the crew should swap SG A Level channel to ____(2)_____ ?

A. (1) NK01 (2) AE LT-519 B. (1) NK01 (2) AE LT-551 C. (1) NG01A (2) AE LT-519 D. (1) NG01A (2) AE LT-551 Answer: A

NRC Written Examination Callaway Plant Reactor Operator Explanation:

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

With the indications of NK11 showing a discharge and bus voltage, with no current flow from the charger NK21, it is indicative of the battery supplying power to NK01 which in turn is supplying power to NN01 due the normal lineup. The indications of NK12 and NK02 are shown for comparison to normal values. AE LT-551 is powered from the NK01 as shown on Step A2. AE LT-519 should be selected.

B. Incorrect. Indications show that NK01 still has bus voltage and therefore has not lost power, (as the same in explanation a) AE LT-551 is powered from the NK01 as shown on Step A2. AE LT-519 should be selected.

C. Incorrect. Plausible because NG01A is the alternate power supply through both the static transfer switch and the SOLA Transformer. In the event NK01 is lost, then NG01A will be supplying NN01 via the static transfer switch. If the static transfer switch is lost then the SOLA transformer can be placed in service to directly supply NN01 via a manual breaker transfer on NN01. AE LT-551 is powered from the NK01 as shown on Step A2. AE LT-519 should be selected.

D. Incorrect. Plausible because NG01A is the alternate power supply through both the static transfer switch and the SOLA Transformer. In the event NK01 is lost, then NG01A will be supplying NN01 via the static transfer switch. If the static transfer switch is lost then the SOLA transformer can be placed in service to directly supply NN01 via a manual breaker transfer on NN01. AE LT-551 is powered from the NK01 as shown on Step A2. AE LT-519 should be selected.

Technical Reference(s):

1. OTO-NK-00001, Failure of NK Battery Charger Rev 14 References to be provided to applicants during examination: None Learning Objective: : T61.003B, Off-Normal Operations, LP#26, Objective B. Describe symptoms or entry conditions for OTO-NK-00001, Failure of NK Battery Charger.

Question Source: Bank # ______

Modified Bank # _X__L17571___

New _______

Question History: Last NRC Exam __2016______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____3_____

10 CFR Part 55 Content:

Comments:

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

Air receivers.

Question #50 Reactor Power is 100%.

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

The leakage exceeds the capacity of the starting air compressors (Assume no other leakage)

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

(1) (2)

A. will are B. will are not C. will not are D. will not are not Answer: B Explanation: The starting air compressor system for the B EDG consists of 2 Starting air tanks TKJ02C and TKJ02D. These are refered to as the C and D starting air receivers. Each Starting air tank has an inlet check valve, KJV711B and KJV712B, that prevents a flaw / depressurization in one air tank from affecting the other tank. The discharge of the starting air tanks are NOT cross connected, i.e. KJV760B is closed.

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

B. Correct - the D Air receiver has sufficient starting air pressure required for more than one start attempt per the Tech Spec basis. Prior to leak on the C Air receiver, it is assumed that starting receiver pressure was 610- 640 psig (normal band) The Air receivers are not cross connected

NRC Written Examination Callaway Plant Reactor Operator during normal operation to ensure redundancy in the starting capability of the EDG. The inlet lines have check valves which allow a single air compressor to supply both receivers simultaneously while still ensuring independence and redundancy of the system.

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

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

Technical Reference(s):

1. M-22KJ02, P&ID Standby Diesel Generator "A" Intake Exhaust, F.O. &Start Air System, Rev 20 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#3 KJ NE Standby Generation Objective C:

Describe the purpose, major components and operation of the following Standby Diesel Generator support systems: Air Start System Question Source: Bank # __X_L16192_

Modified Bank # ______

New _______

Question History: Last NRC Exam __2019 NRC Exam __________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Process Radiation Monitoring Group # 1 K/A # 073 K4.01 Importance Rating 4.0 Knowledge of PRM system design feature(s) and/or interlock(s) which provide for the following:

Release termination when radiation exceeds setpoint Question #51 A Waste Gas Decay Tank release is in progress when the following annunciators alarm:

61A, Process Rad HIHI

61B, Process Rad HI The BOP informs the CRS that GH-RE-10B, Radwaste Bldg Exh Rad Monitor, indicates a Red alarm.

What automatic action(s) should occur due to this condition?

A. HA HCV-14, Radiation Control Valve, closes ONLY B. HA HCV-14, Radiation Control Valve, closes and Radwaste Building Exhaust Fans, CGH01A/B, trip ONLY C. HA HCV-14, Radiation Control Valve, closes; Radwaste Building Exhaust Fans, CGH01A/B, trip; and Control Room Ventilation Isolation (CRVIS) ONLY D. HA HCV-14, Radiation Control Valve, closes; Radwaste Building Exhaust Fans, CGH01A/B, trip; Control Room Ventilation Isolation (CRVIS); and Containment Purge Isolation (CPIS)

Answer: A Explanation:

Annunciator 61A, refers the crew to OTA-SP-RM011 to determine which detector is in alarm. 1 shows that A Hi-Hi radiation alarm on GH RE-10B will close HCV-14 to terminate the release.

A. Correct B. Incorrect - while HA HCV14 closes the Radwaste Building Exhaust damper dont trip.

Plausible as other ventilation system exhaust fans trip when a rad monitor indicates a release is in progress ex. CPIS and CRVIS have exhaust fans that trip on an isolation signal.

NRC Written Examination Callaway Plant Reactor Operator C. Incorrect - See above explanation. CRVIS in addition is plausible as 6 ventilation rad monitors will generate a CRVIS GG-RE 27&28, CK-RE-04/05, and GT-RE-22/33 D. Incorrect - See above explanation. Plausible as stated above and it is believe that GH-RE-10B functions the same way as GT-RE22/33 which provide a CRVIS and CPIS when tripped.

This distractor basically isolates all plant ventilation which is a plausible option when dealing with a gaseous release.

Technical Reference(s):

1. OTA-SP-RM011, Radiation Monitor Control Panel RM-11, Rev 44 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#16, Objective B: DESCRIBE the purpose and operation of the following Gaseous Radwaste System components:

1. Decay Tanks

2. Waste Gas Tank Sampling / Monitoring

3. Waste Gas release and isolation

4. Radiation Monitoring Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____4____

10 CFR Part 55 Content:

Comments:

To prevent excessive focus on the liquid discharges for SRO candidates, this question is focused on the gaseous system and associated interlocks

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Service Water Group # 1 K/A # 076 A2.01 Importance Rating 3.5 Ability to (a) predict the impacts of the following malfunctions or operations on the SWS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of SWS.

Question #52 Reactor power is 100%:

Annunciator 12A, SERV WTR PMP LOCKOUT, is LIT

Annunciator 12D, SERV WTR PMP TROUBLE, is LIT The local Operator reports local alarm 1B, SW STRAINER TROUBLE, is LIT and strainer differential is 3 psid.

What condition could cause the local alarm and what action is required?

A. Auto start of a Service Water pump with its strainer differential pressure switch equalizing valve shut, ensure the discharge valve is closed on any tripped Service Water pump B. Auto start of a Service Water pump with its strainer differential pressure switch equalizing valve open, ensure the discharge valve opened on pump that auto started C. Trip of a running pump and failure to auto start the standby pump, start a standby Service Water pump D. Trip of a running Service water pump and ESW pump is running, start a standby Service Water pump Answer: A Explanation:

OTN-EA-00001 contains a NOTE before step 5.2.6 that states: WHEN a service water pump auto starts, Annunciator 1B, SW STRAINER TROUBLE, (on CPEA2105, ANNUNCIATOR AND CONTROL PANEL, SERV WATER SYSTEM), is an expected alarm, due to air trapped in the sensing line." The normal alignment has the equalizing valves shut on the standby pump per step 5.3.2.c. OTA-RK-00014 Addendum 12A , SERV WTR PMP LOCKOUT, step 1.1 AUTO ACTION states: Standby Pump AUTO starts. Section 3.0 OPERATOR ACTIONS step 3.4 states:

ENSURE the discharge valve on the tripped pump is CLOSED. Step 3.6 states: IF a Standby pump was started, refer to OTN-EA-00001.

NRC Written Examination Callaway Plant Reactor Operator A. Correct see above B. Incorrect - Auto start of a Service Water pump with its VEA2168 valve open, open the discharge valve on pump that auto started - Plausible since the alarm is expected on an auto start, however the VES2168 valve being open would prevent the local alarm and the discharge valve auto opens.

C. Incorrect -Trip of a running pump and failure to auto start the standby pump, start a standby Service Water pump - Plausible since OTA-RK-00014 Addendum 12A step 3.1 states: IF Standby Service Water Pump did NOT AUTO start, START Standby Service Water Pump, however the local alarm is caused by the auto start of the standby pump in normal alignment.

D. Incorrect - Trip of a running Service water pump and ESW pump is running, start a standby Service Water pump - Plausible since the OTA-RK-00014 Addendum 12A provides the guidance, however it fails to address the local alarm, expected for plant alignment within the normal operating band.

Technical Reference(s):

1. OTA-RK-00014, Addendum 12A, Rev 9

2. OTN-EA-00001, Service Water System, Rev 41 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP-4, CIRC WATER SYSTEM (DA) SERVICE WATER SYSTEM (EA) CWCCS (DD), Obj. F. Describe the basic Service Water flow paths

1. Service Water

2. Service to Essential Service Water Question Source: Bank # _X__L23217_____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____4____

10 CFR Part 55 Content:

Comments:

Verified no overlap with Audit Question#51.

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Service Water Group # 1 K/A # 076 K1.21 Importance Rating 2.7 Knowledge of the physical connections and/or cause- effect relationships between the SWS and the following systems: Auxiliary backup SWS Question #53 Reactor Power is 100%.

2 Service Water Pumps are running

An NB02 undervoltage condition occurs What is the status of and actions required for the Service Water system?

A. Service Water will be supplying Turbine building loads ONLY. Secure one service water pump within 5 minutes.

B. Service Water will be supplying Turbine building loads and "A" Train ESW loads.

Secure one service water pump within 5 minutes.

C. Service Water will be supplying Turbine building loads ONLY. Open the Cooling Tower Bypass Valves to lower service water pressure.

D. Service Water will be supplying Turbine building loads and "A" Train ESW loads.

Open the Cooling Tower Bypass Valves to lower service water pressure.

Answer: A Explanation: As a result of the undervoltage, the B EDG will start and close in on the NB02 which will start the shutdown sequencer which will start the B ESW pump. This will cause ESW to isolate from service water as both the supply and return cross connects will close. Note: Each train has 2 cross connects on the supply and the return side, one of the supply and one return are opposite train powered and logic controlled. This in turn would start the opposite train ESW pump due to undervoltage on the opposite train bus (NB02) with low flow through the A trains containment cooler. So the final configuration would be both ESW pumps running each supplying their own trains loads.

While it is listed in both OTN-EA-00001 and OTN-EF-00001, a caution in OTN-EF-00001 explains it the best " If this will be the second ESW pump started and two SW pumps will be supplying only the Turbine Building loads, a SW pump will have to be secured within five minutes of ESW Pump start."

NRC Written Examination Callaway Plant Reactor Operator The distractor of opening the cooling tower bypass valve is plausible as the Service water and Circ Water lines come together and return to the cooling tower and if the candidate believes there is procedural direction to open this valve to divert service water to the cooling tower and hence lower service water pressure, it is plausible. The bypass valve is only done for cold weather operation to raise cooling tower basin temperature per OTN-DA-00001 ADD 4.

The distractor of SW still supplying A Train ESW load is plausible this is the normal flowpath and the stem does not provide any A Train cues; i.e. the candidate does not understand or remember the cross train powered valves for the supply or return cross connects nor the ESW pumps auto start on opposite train UV and low flow (reference OTA-RK-00020, ADD 54A)

A. Correct - See above explanation B. Incorrect - this is the wrong flowpath C. Incorrect - wrong action as securing one of the SW pumps is required D. Incorrect - both are wrong Technical Reference(s):

1. OTN-EA-00001, Service Water System, Rev 41

2. OTN-EF-00001, Essential Service Water, Rev 79

3. M-22EF01, P&ID, ESW, Rev 82

4. M-22EF02, P&ID, ESW, Rev 78

5. OTA-RK-00020, Addendum 54A, ESW A Pressure Low / Flow Low, Rev 3

6. OTN-DA-00001, Addendum 4, Cooling Tower Operation, Rev 14 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP #4, Objective F & G:

F: LIST the systems that interface with the Circulating and Service Water Systems and EXPLAIN how a loss of the interfacing system or a loss of the Circulating or Service Water Systems affects the other.

G.DESCRIBE the response of the Service Water System to a Safety Injection Signal, Loss of Offsite Power, Aux Feed Low Suction Pressure.

Question Source: Bank # _X___No ID_____

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2016_____

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____3_____

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

Comments:

In terms of this K/A, the Service Water (SW) systems is the SWS and the ESW system is the auxiliary backup SWS.

The undervoltage condition in the question stem is merely for the auto start of the ESW pump in the same train ( B Train) and provides half of the A Trains ESW start signal (low flow still needed but will occur when ESW trains disconnect from SW). The UV sets up the cause and effect situation between SW and ESW i.e. what is the status of the SW system when the auxiliary SW train(s) receive an autostart signal.

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Instrument Air Group # 1 K/A # 078 K3.02 Importance Rating 3.4 Knowledge of the effect that a loss or malfunction of the IAS will have on the following:

Systems having pneumatic valves and controls Question #54 A loss of offsite power occurred.

15 minutes later, a Safety Injection occurred due to a faulted SG.

EFHV0043 and EFHV0044, ESW to 'A' and 'B' air compressors, are closed.

What caused EFHV0043 and EFHV0044 to close?

A. LOCA Load Shed B. Blackout Load Shed C. Loss of Instrument Air D. High DP Across valves Answer: C Explanation:

A. Incorrect - plausible as several sets of EF valves will reposition on either a SIS or LOOP Load Sheds ex per M-22EF01 EFHV0023 and EFHV0025 closes on either B. Incorrect - plausible for same reason as choice A C. Correct - per M-22EF02, EFHV0043/44 fail closed on a loss of air.

D. Incorrect - plausible as per M-22EF02 there are PDT (differential pressure transmitters) around EFHF43 and 44 and it may be believed this generates a signal to close the values to protect / isolate leaks downstream of the valve.

Technical Reference(s):

1. OTO-KA-00001, Partial or total Loss of Instrument Air, Rev 29

2. M-22EF01, P&ID ESW System - Sheet 1, Rev 82

3. M-22EF02, P&ID ESW System - Sheet 2, Rev 78 References to be provided to applicants during examination: None

NRC Written Examination Callaway Plant Reactor Operator Learning Objective: T61.0110, Systems, LP#14, Objective H: LIST the major interface systems with the Service and Instrument Air System and EXPLAIN how a loss of the interfacing system or a loss of the Service and Instrument Air System affects the other.

Question Source: Bank # _X R14232_____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

LOD ____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Containment Group # 1 K/A # 103 A4.04 Importance Rating 3.5 Ability to manually operate and/or monitor in the control room: Phase A and Phase B resets Question #55 Following a large break LOCA that resulted in a Containment Spray Signal (CSAS) and Containment Isolation Signal - Phase B (CIS-B), what are the MINIMUM actions required to reset the Containment Isolation Signal - Phase B actuation? (Note: Assume all CIS-B and CSAS signals are clear.)

A. Depress both SB HS-51 & SB HS-54, CSAS reset pushbuttons ONLY B. Depress both SB HS-52 & SB HS-55, CIS-B reset pushbuttons ONLY C. Depress both SB HS-52 & SB HS-55, CIS-B reset pushbuttons, and depress both SB HS-54 and SB HS-51, CSAS reset pushbuttons D. Depress both SB HS-53 & SB HS-56, CIS-A (Containment Isolation Signal-Phase A) reset pushbuttons, and depress both SB HS-52 & SB HS-55, CIS-B reset pushbuttons Answer: B Explanation:

Containment Isolation Signal Phase A (CISA), which functions to isolate all lines penetrating containment that are not essential to reactor protection in order to prevent the release of radioactivity to the environment. Any of the following can generate a CISA:

Manual, either of two switches @ RL018.

SIS.

A CISA signal can be reset at any time following its actuation without having to reset any other signal.

Containment Isolation Signal Phase B (CISB), which completes isolation of all lines penetrating containment not essential for reactor protection. Any of the following can generate a CISB:

Manual, two switches (four total) @ RL018.

Containment High 3 pressure, 27 psig, 2/4 channels.

Per OTO-SA-00001, Attachment D, CIS-B recovery, Step D1 and D2, to reset the CIS-B:

"D2. RESET CIS-B (RL018):

SB HS-52

SB HS-55 Additionally, ES-1.2 Step 2 states: "RESET Containment Isolation Phase A And Phase B:

NRC Written Examination Callaway Plant Reactor Operator

Phase A (CISA):

o SB HS-53 o SB HS-56

Phase B (CISB):

o SB HS-52 o SB HS-55" Therefore, to reset phase B containment isolations, the Minimum required actions is to Depress both SB HS-52 & SB HS-55, CIS-B reset pushbuttons ONLY.

The Note "assuming all signals are clear" was incorporated such that it can't be argued that no above action would be successful.

A. Incorrect -Plausible if it is believed that the resetting of the CSAS is the same as resetting CIS-B since they are actuated on the same setpoint. Resetting CSAS by depressing these PB is step E2 of OTO-SA-00001 B. Correct - See above explanations C. Incorrect - Plausible if it is believed that the CSAS must be reset in conjunction with the CIS-B as they both have the same actuation setpoints.

D. Incorrect - Plausible if it is believed that the phase A (which occurs at CTMT Hi 1 - 3.5 psig) must be reset prior to or in conjunction with resetting the CIS-B (which occurs at CTMT Hi 3) and is procedure step ES-1.2.

Technical Reference(s):

1. OTA-SA-00001, EFSAS Verification and Restoration, Rev 42 - Attachments C,D,E & AO

2. ES-1.2, Post LOCA Cooldown and Depressurization, Rev 20 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#27, Objective F & K:

F. LIST the Engineered Safety Features (ESF) actuations that may be generated by the RPS.

K. IDENTIFY the RPS Main Control Board controls, alarms and indications and EXPLAIN how each is used to predict, monitor or control the RPS.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Control Rod Drive Group # 2 K/A # 001 K6.11 Importance Rating 2.9 Knowledge of the effect of a loss or malfunction on the following CRDS components: Location and operation of CRDS fault detection (trouble alarms) and reset system, including rod control annunciator Question #56 The crew is raising power to full power after a refueling outage when:

Ann 79A, ROD CTRL URG Fail, alarms

It has been determined that a Power Cabinet Urgent Failure alarm caused by a Regulation Failure occurred (1) What is the impact on the Rod Control System?

And (2) What is required to reset and restore the Rod Control System?

A. (1) ALL rod motion is inhibited (2) The crew needs to reset the alarm only B. (1) ALL rod motion is inhibited (2) I&C troubleshooting and associated actions are required prior to crew resetting the alarm C. (1) Rod motion is ONLY inhibited in the affected power cabinet (2) The crew needs to reset the alarm only D. (1) Rod motion is ONLY inhibited in the affected power cabinet (2) I&C troubleshooting and associated actions are required prior to crew resettng the alarm Answer: D Explanation:

Per Annun 79A,

A Logic Cabinet Urgent Failure prevents all rod motion in manual and automatic.

A Power Cabinet Urgent Failure prevents rod motion in manual and automatic for rods in the affected power cabinet.

NRC Written Examination Callaway Plant Reactor Operator And per the table in the annunciator, a regulation failure (where coil current is not equal to the demand) is a power cabinet urgent failure which prevents rod motion in the effected group only.

If it is believed a regulation failure is a logic cabinet failure then inhibiting all rod motion would be correct, making it a plausible choice.

Per OTO-SF-00001 step #12 states "DIRECT I&C To Perform ITM-ZZ-0015, Rod Control Troubleshooting Guidelines. As the power and logic cabinets are on the back panels in the control room, it is plausible that a regulation failure may be reset at the cabinet (similar to reset a flag on a breaker) and rod control restored without I&C support.

A. Incorrect - both are wrong, See above explanation B. Incorrect - part a is wrong, See above explanation C. Incorrect - part b is wrong. See above explanation D. Correct Technical Reference(s):

1. OTO-SF-00001, Rod Control Malfunctions, Rev 21

2. OTA-RK-00022, Annunciator 79A, Rod Control Urgent Failure, Rev 002 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#26, Objective M: LIST the causes of a Rod Control Urgent Failure and EXPLAIN the effects on the system.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Rod Position Indication Group # 2 K/A # 014 K4.06 Importance Rating 3.4 Knowledge of RPIS design feature(s) and/or interlock(s) which provide for the following: Individual and group misalignment Question #57 The crew is performing OTG-ZZ-00002, Reactor Startup. The procedure directs you to verify proper bank overlap as rods are withdrawn.

What rod positions represents proper Control Bank Overlap?

A. Bank B at 115 steps and Bank C at 3 steps B. Bank B at 200 steps and Bank C at 87 steps C. Bank A at 220 steps and Bank B at 105 steps D. Bank A at 228 steps and Bank B at 110 steps Answer: C Explanation:

A. Incorrect-Bank C should be at 85 steps B. Incorrect-Bank C should be at 0 steps C Correct-Bank B begins stepping out when Bank A reaches 115 steps.

D. Incorrect-Bank B should be at 113 steps Technical Reference(s):

1. TS 3.1.6, control Bank Insertion Limits

2. Curve Book Figure 13-1, COLR Section 2.4.3, Rev 061 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP# Rod control, Enabling Objective J, LIST and DESCRIBE the control interlocks associated with rod control.

Question Source: Bank # _X___L5183_____

NRC Written Examination Callaway Plant Reactor Operator Modified Bank # ______

New _______

Question History: Last NRC Exam ___2004______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____2_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Nuclear Instrumentation Group # 2 K/A # 015 K2.01 Importance Rating 3.3 Knowledge of bus power supplies to the following: NIS channels, components, and interconnections Question #58 What is the power supply to Intermediate Range Nuclear Instrument N35?

A. NN01 B. NN02 C. NN03 D. NN04 Answer: A Explanation:

NN01 supplies SR N31, IR N35, PR N41 instruments NN02 supplies SR N32, IR N36, PR N42 instruments NN03 supplies PR N43 instrument NN04 supplies PR N44 plus other various Comparator and rate and audio count rate instrumentation.

Gamma metrics NI (SEN 60/61) are supplied as follows:

NN01 feeds SEN60 detector, preamplifier, signal processor and meters.

NN04 feeds SEN61 detector, preamplifier, signal processor, meters and recorder.

All other 120VAC NN buses are plausible as they are all safety related 120 VAC NN buses and supply at least one Nuclear Instrument subsystem as listed above.

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

1. E-23NN01, Class 1E Instrument AC Schematic, Rev 14

2. E-23SE02(Q), Ex-core Neutron Monitoring System Protection Rack SE054B, Rev 6

NRC Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems, LP #28, Excore Nuclear Instrumentation, Objective B:

DESCRIBE the purpose and operation of the following Nuclear Instrumentation System (NIS) components:

1. Source Range Nuclear Instruments

2. Intermediate Range Nuclear Instruments

3. Power Range Nuclear Instruments Question Source: Bank # ______

Modified Bank # ______

New __X_____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____2____

10 CFR Part 55 Content:

Comments:

This question was compared to question 40 and the use of the same possible choices (120VAC safety related buses) in each question is sound. The same possible choices but different answers prevents overlap and one question helping answer another question.

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 In Core Temperature Monitor Group # 2 K/A # 017 K6.01 Importance Rating 2.7 Knowledge of the effect of a loss or malfunction of the following ITM system components:

Sensors and detectors Question #59 (1) How will the associated RCS Degrees Subcooling channel respond to a Core Exit Thermocouple (CET) with an OPEN circuit?

And (2) What is the MINIMUM operability requirement of Core Exit Thermocouples in accordance with Technical Specifications?

A. (1) No change (2) 1 channel per quadrant B. (1) No change (2) 2 channels per quadrant C. (1) Superheated (2) 1 channel per quadrant D. (1) Superheated (2) 2 channels per quadrant Answer: B Explanation:

Per TS 3.3.3, 2 channels per quadrant are required. 1 channel is plausible as a channel consists of 2 CETs. If it is believed that 2 CETS are required per quadrant (vice 2 required channels) the by applying the number of CETs in a channel, 1 channel would be selected.

Each RCS subcooling microprocessors receives input data from:

Two wide range hot leg temperature RTD's.

Two wide range cold leg temperature RTD's.

One RCS wide range pressure transmitter (PT 403 or 405)

Twenty five core thermocouples (T/C's).

Three thermocouple reference junction box temperature RTD's.

Two PZR narrow range pressure transmitters.

NRC Written Examination Callaway Plant Reactor Operator Each microprocessor calculates saturation temperature based on auctioneered low RCS and Pressurizer pressure signals. Using the calculated saturation temperature, the following information is then generated:

Margin to saturation (subcooling); based on hottest RCS loop RTD.

Margin to saturation (subcooling); based on hottest core T/C.

On a CET open circuit, the output fails low. As the subcooling monitor uses the highest CET, the calculation will either remain unchanged or (if it was the highest CET) show slightly more subcooling. Superheated is plausible if the failure of the open circuit isnt understood (i.e. fails high) or the temperatures in the subcooling monitor are crossed (i.e. CET temperature indication goes down lowering below the saturation temperature value)

A. Incorrect - wrong TS information B. Correct C. Incorrect - both subparts are wrong D. Incorrect - superheated determination is wrong as explained above Technical Reference(s):

1. Technical Specifications 3.3.3, PAM Instrumentation References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#29, Objective G & K:

G. DESCRIBE the function of the core subcooling monitor.

K: DESCRIBE failure modes and resulting indication for incore thermocouples.

Question Source: Bank # _X__L22959____

Modified Bank # ______

New _______

LOD ____3____

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Spent Fuel Cooling Group # 2 K/A # 033 A1.01 Importance Rating 2.7 Ability to predict and/or monitor changes in parameters (to prevent exceeding design limits) associated with Spent Fuel Pool Cooling System operating the controls including: Spent fuel pool water level Question #60 Given the following plant conditions:

CRDM latch and drag testing is being performed following core load

ECV0995, Fuel Transfer Tube Isolation Valve is open

A cavity seal ring leak occurs

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

Spent Fuel Pool Level indicates -24 inches What pump(s) should have tripped?

A. Fuel Pool Skimmer pump B. Fuel Pool Cleanup pumps C. Residual Heat Removal pumps D. Spent Fuel Pool Cooling pumps Answer: D Explanation:

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

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

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

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

1. OTN EC 00001, Fuel Pool Cooling and Cleanup System, Rev 44

2. OOA-BB-00003, Refueling Level Indications, Rev 14 References to be provided to applicants during examination: None Learning Objective: T61.0110.6 Systems LP #24, Objective E - EXPLAIN the design features that prevent draining the Spent Fuel Pool.

Question Source: Bank # __X___L12185____

Modified Bank # ______

New _______

Question History: Last NRC Exam __2014_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3______

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Steam Generator Group # 2 K/A # 035 K5.03 Importance Rating 2.8 Knowledge of operational implications of the following concepts as the apply to the S/GS: Shrink and swell concept Question #61 Compare two steady state reactors: 'A' Reactor at 40% power and 'B' Reactor at 20% power.

(1) If a reactor and turbine trip occur, SG levels will shrink LESS in Reactor .

And (2) If either reactor has a stabilized idle RCS Loop when a reactor and turbine trip occur, SG level in the idle loop will be ____(2)_____ than in the remaining active loops immediately following the reactor/turbine trip.

A. (1) 'A' (2) Lower B. (1) 'A' (2) Higher C. (1) 'B' (2) Lower D. (1) 'B' (2) Higher Answer: D Explanation:

The reactor with a higher power level and hence higher rate of heat transfer in the SG has more voiding and upon a reactor and turbine trip, will shrink more than a reactor at lower power.

With an idle loop that is giving time to stabilize and restore its SG level, and when a reactor/

turbine trip occurs there is less initial heat transfer through that loop which means less voiding and less shrink upon a trip. Therefore, the idle loop SG level will be higher as compared to the active loop immediately following a reactor / turbine trip.

NRC Written Examination Callaway Plant Reactor Operator The opposite cases are presented as distractors and is plausible if the candidate incorrectly models the secondary heat sink and its response when the amount of heat input from the RCS drastically decreases.

A. Incorrect - see above explanations - both parts are wrong B. Incorrect - see above explanations - part 1 is wrong C. Incorrect - see above explanations - part 2 is wrong D. Correct Technical Reference(s):

1. UFSAR Section 5.4.2 Steam Generators

2. OTG-ZZ-00004, Power Operation, Rev 99 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#20, Objective B: DESCRIBE the purpose and operation of the following Main Steam System components:

6. Steam Generator Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ___2____

10 CFR Part 55 Content:

Comments:

k/a match as there isnt a plausible situation in which there would be a SG level swell without entering the major accident category. The shrink topic was examined in part 1 and part 2 of the question examines when SG level would be higher than others effective testing a SG level swell concept.

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Steam Dump/Turbine Bypass Control Group # 2 K/A # 041 K4.18 Importance Rating 3.4 Knowledge of SDS design feature(s) and/or interlock(s) which provide for the following: Turbine trip Question #62 The Main Turbine tripped from 100% power.

1 minute after the turbine trip Tavg is 580°F.

What is the expected position of the condenser steam dump valves?

Full Open Modulating A. 3 9 B. 6 6 C. 9 3 D. 12 0 Answer: C Explanation:

The 12 Condenser Steam Dumps are broken down into 4 groups. When Power is greater than 15%, the steam dump controller is in the Tavg mode of operation and 2 options are available:

load reject and plant trip modes.

The Tavg mode (load rejection) will reduce Tavg to its programmed value following a load rejection when the transient is greater than the capacity of the rod control system. The Tavg mode (plant trip) will reduce Tavg to the no load Tavg reference setpoint (557*F) following a reactor trip.

On a plant trip from 100 %, where Tavg was ~585 F, the heat from the nuclear fission process stops, and the control circuitry returns main steam to the mode three setting (557 F) to stabilize the plant without the need for MSSVs or ASD operation.

Upon initiation of condenser steam dump operations, Group No. 1 modulates open. Group No. 2 does not begin to modulate open until Group No. 1 is fully open. This progression continues until all four groups are open. The dump valves close in the reverse sequence such that Group No. 4 fully closes prior to Group No. 3 starting to close and so on. The Group No. 1 dump valves are used as the cooldown valves for plant cooldown below Tavg of 550*F.

In the Tavg mode (plant trip), the positioning load signal to the I/P converter varies with Tavg -

Tno-load difference. No-load Tavg (Tno-load) is 557*F. The (compensated) auctioneered high

NRC Written Examination Callaway Plant Reactor Operator Tavg signal and Tno-load signal are compared, producing a signal proportional to the Tavg - Tno-load difference. This deviation determines the positioning signal generated by the plant trip controller. In the Tavg mode (plant trip), the signal is ~3.5% /F DT.

The Tavg mode (plant trip) also has high-1 and high-2 trip bistables, which will act in the same manner as for the Tavg mode (load rejection) trip bistables. The actual bistable trip setpoints will be different due to the different percent power/F DT ratios. Hi-1 setpoint is 14.15F DT. The Hi-2 setpoint is 28.3F DT. When the Hgh-1 bistable trips, the trip solenoid valves (No. 4) for Groups 1 and 2 dump valves energize, fully opening Group 1 and 2 dump valves. Similar for High-2 bistables and Groups 3 and 4 dump valves Tavg - Tref for the stem conditions is 580 - 557F = 23F. Using the 3.53% per F, this equals a 81.2% which means that groups 1 - 3 are full open with group 4 partial open (or modulating). 1 minute after the trip is adequate time for the High 2 bistable to reset A. Incorrect - Plausible as group 1 are the cooldown valves used to cooldown the plant below 550F and the applicant believes these valves are full open with all other valves partially open due to the Tavg - Tref difference. This distractor applies the concept of all steam dumps valves (except the cooldown group) module together just like the main turbine control valves.

B. Incorrect - Plausible as the high1 setpoint is met and if it was believed that 3 and 4 modulate together. Until temperature lowers below the high 1 setpoint.

C. Correct D. Incorrect - plausible if the high 2 setpoint from the load reject (14.15F) was remember or if it is believed that all 4 groups will be full open until the high 1 setpoint is reached or not enough time or temperature decrease has occurred for the real high 2 setpoint (28.3F) to be cleared.

Technical Reference(s):

1. 7250D64 Sheet 10, Functional Diagram Steam Dump Control, Rev 4 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#20, Objective B & J:

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

2. Main Steam Isolation and Bypass Valves

3. MSL Low Point Drains

4. Steam Dumps and Isolation Valves J. Given specific plant conditions, CALCULATE and DETERMINE the expected position of the Steam Dumps.

Question Source: Bank # ______

Modified Bank # __X L23397____

New _______

Question History: Last NRC Exam ____N/A________

NRC Written Examination Callaway Plant Reactor Operator Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____3_____

10 CFR Part 55 Content:

Comments:

The position of full closed was removed from the stem as it provided 3 choices which made for easy distractor elimination

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Condenser Air Removal Group # 2 K/A # 055 K1.06 Importance Rating 2.6 Knowledge of the physical connections and/or cause-effect relationships between the CARS and the following systems: PRM system Question #63 With the plant at full power, AB-RE-16A-D, Main Steam Line Monitors (N16) have been declared inoperable due to common mode failure What radiation monitor serves as the primary backup to satisfy the Primary to Secondary Leak Rate Program per OTA-SP-RM011?

A. GE-RE-92, Condenser Air Discharge Monitor B. BM-RE-25, Steam Generator Blowdown Processing System Monitor C. BM-RE-52, S/G Blowdown Discharge Pumps Discharge Radiation Element D. SJ-RE-02, Steam Generator Liquid Radioactivity Monitor Answer: A Explanation:

GE-RE-92 is the Condenser Air Removal PRM. Condenser Air discharge monitor, GE-RE-92, actions are outlined in Attachment 9 of OTA-SP-RM011. There are no automatic or immediate actins associated with GE-RE-92. Per Attachment 2 of OTA-SP-RM011, need to check if GE-RE-92 is in operation if N16 monitors fail.

Per APA-ZZ-01023, step 4.1.1. USE Main Steam N-16 PSLR monitors (N-16 monitors) as the preferred primary to secondary leak indication. Per step 4.1.2, IF N-16 monitors can NOT be used as primary to secondary leak indication AND IF GERE0092 is OPERABLE, USE GERE0092 as primary to secondary leak indication instead.

The distractors are from step #3.a and a note prior to step 3.d. Step 3.a list PRMs which are used to the alarm and the note identifies secondary side monitors which would show a higher RCS activity.

Step 3.a - CONFIRM the alarm using the following:

BM-RE-25

BM-RE-52 NOTE - Increasing RCS activity can cause increases on secondary side activity monitors GE-RE-92, SJ-RE-02 and BM-RE-25.

NRC Written Examination Callaway Plant Reactor Operator A. Correct B. Incorrect - See above explanation C. Incorrect - See above explanation D. Incorrect - See above explanation Technical Reference(s):

1. OTA-RM-SP011, Radiation Monitor Control Panel RM-11, Rev 44

2. APA-ZZ-01023, Primary to Secondary Leakage Program, Rev 027 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#22, Objective C & K & L C. DESCRIBE the function and operation of the following Condensate System components:

9. Condenser Air Removal Subsystem

a. Condenser Vacuum Pumps K. STATE the function and EXPLAIN the design criteria of the Condenser Air Removal System.

L. LIST the automatic action setpoints for the Condenser Air Removal System.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X__

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

10 CFR 55.41(b)(4)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Liquid Radwaste Group # 2 K/A # 068 K1.07 Importance Rating 2.7 Knowledge of the physical connections and/or cause effect relationships between the Liquid Radwaste System and the following systems: Sources of liquid wastes for LRS Question #64 Which one of the following is a direct input into the Reactor Coolant Drain Tank (RCDT)?

A. RCP #2 Seal Leakoff B. RCP #3 Seal Leakoff C. RHR Suction Relief(s), EJ-8708A/B D. RCP Seal Water Return Relief, BG-8121 Answer: A Explanation:

The Liquid Radwaste system consists of basic waste streams, segregated to prevent intermixing.

The reactor coolant drain tank (RCDT) and equipment drains process the water originally meant to be recycled back to the plant. Water originally meant to be discharged is processed by floor drains. Laundry water, containing detergent, is also segregated for processing separately.

The RCDT influents include:

RCP number two seal leakoff

Excess letdown

Reactor vessel flange leakoff

ECCS accumulator drains

RCS loop drains

PRT drainage

Refueling pool drains

Miscellaneous valve stem leakoffs A. Correct B. Incorrect - Leakoff from the #3 seal is directed to the containment sump C. Incorrect - Per print M-22BB02, RHR suction relief valves discharge into the PRT D. Incorrect - Per print M-22BB02, Reactor Coolant Pump Seal Return Relief (BG-8121) is direct to the PRT. This is also shown on print M-22BG01

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

1. M-22HB01, P&ID Liquid Radwaste System - Sheet 1, Rev 39

2. M-22BG01, P&ID CVCS, Rev 33

3. M-22BB02, P&ID RCS, Rev 33 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP #16, Objective F: LIST the systems that interface with the Liquid Radwaste system and IDENTIFY the systems which send influent to the Liquid Radwaste system.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

LOD ___3____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 2 Circulating Water Group # 2 K/A # 075 A2.01 Importance Rating 3.0 Ability to (a) predict the impacts of the following malfunctions or operations on the circulating water system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of intake structure Question #65 With Reactor Power at 80%, security reports significant storm damage to the Circ and Service Water Building.

Current plant conditions are:

Annunciators 12A, Serv Wtr Pmp Lockout, and 13A, Circ WTR PMP Lockout, are LIT

2 Circ Water Pumps are tripped

1 Service Water pump is tripped

Condenser Vacuum is 5.3" HgA and rising

Stator Cooling outlet temperature, (CETI 38A), is 70°C and rising What is the NEXT action the crew should take?

A. Trip main turbine and proceed to OTO-AC-00001, Turbine Trip Below P-9 B. Check Turbine Runback is in progress and refer to OTO-MA-00001, Turbine Load Rejection C. Check Turbine Setback is in progress and refer to OTO-AD-00001, Loss of Condenser Vacuum D. Manually start both ESW trains and isolate from Service Water then refer to OTN-EF-00001, ESW System Answer: C Explanation:

With the conditions given in the stem, 2 circ water pumps and 1 service water pump tripped but the standby pump autostarted, leaving the same number of service water pumps in service as before the damage.

The distractor of "Manually start both ESW trains and isolate from Service Water then refer to OTN-EF-00001, ESW System" is from OTA-RK-00014 Addendum 12A, Service Water 3.3. IF less than two Service Water pumps are running, START both ESW trains and isolate from

NRC Written Examination Callaway Plant Reactor Operator Service Water ." This is a plausible distractor as the stem indicates that 1 service water pump has tripped and if the autostart of the standby pump is not properly applied then this action is a correct future action.

The distractor of "Trip the main turbine and proceed to OTO-AC-00001, Turbine trip Below P-9 is plausible due to condenser vacuum rising with the automatic turbine trip @8.5"HgA.

The correct answer of "Check Turbine Setback is in progress and refer to OTO-AD-00001, Loss of Condenser Vacuum" is from Annunciator 13A step 3.3 that states: IF ALL of the following conditions exist, CHECK Turbine Setback is in progress:

An operating Circ Water Pump has tripped

Turbine Load is greater than 75%

DAHS0113, CIRC WATER PUMP TURB SETBACK, is in ENABLE

Condenser Vacuum is greater than 5.0 HgA

Annunciator 117B, SETBACK/RUNBACK ACTIVE, is lit And step 3.4 that states" Refer To OTO-AD-00001, Loss of Condenser Vacuum." Note:

DAHS0113, CIRC WATER PUMP TURB SETBACK, is currently maintained in the ENABLE position.

The distractor of "Check Turbine Runback is in progress and refer to OTO-MA-00001, Turbine Load Rejection" is from annunciator 12A step 3.6.2 which states " IF Stator Cooling outlet temperature Setpoint is met, Go To OTO-MA-00001, Turbine Load Rejection." The value of 70C was chosen such that it is higher than the Gas temperature limit but below stator limit of 82C making the choice incorrect but plausible if the setpoints of parameters incorrectly applied to one another.

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

1. OTA-RK-00014, Addendum 13A, Circulating Water Pump Lockout, Rev 1

2. OTA-RK-00014, Addendum 12A, Service Water Pump Lockout, Rev 9

3. OTO-AD-00001, Loss of Condenser Vacuum. Rev 035 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#4, Objective B. IDENTIFY the Circulating and Service Water System Main Control Board (MCB) controls, alarms, and indications and DESCRIBE how each is used to predict, monitor, or control changes in the Circulating and Service Water Systems.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

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

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ___3____

10 CFR Part 55 Content:

10 CFR 55.41(b)(4)

Comments:

For Callaway, the intake structure is on the Missouri River ~2 miles from the plant and its own system - River Intake (DE) System. A loss of this structure has no short term consequences as the water makeup facility will continue to gravity feed the cooling tower. For this question, intake structure as it pertains to Circulating Water is the building at the base of the cooling tower that contains the Circulating Water and Service Water pumps.

The Question is written to the B part of the k/a which is allowed per Nureg 1021 Rev 11 and the predictive nature is present in the question by asking "what is the next action the crew should take".

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Conduct of Operations Group # N/A K/A # G2.1.26 Importance Rating 3.4 Knowledge of industrial safety procedures (such as rotating equipment, electrical, high temperature, high pressure, caustic, chlorine, oxygen and hydrogen).

Question #66 During rounds in the Demin Building, a small leak from a tank is identified.

Which quadrant of the HAZMAT label (shown below) would show the Health Hazard of the compound inside the tank.

A. White B. Yellow C. Red D. Blue Answer: D Explanation:

NRC Written Examination Callaway Plant Reactor Operator A. Incorrect - this is the special hazard section B. Incorrect - this is the reactivity hazard section C. Incorrect - this is the fire hazard section D. Correct - this is the health hazard section Technical Reference(s):

1. APA-ZZ-00831, Hazardous Chemical Control Program, Rev 20 - Attachment 2 References to be provided to applicants during examination: None Learning Objective: none specific to the ILT program - Chemical Awareness training is provided for all employees with unescorted access and is required for all operations and tracked under job duty code PAT-HAZ by completing annual training/retraining course T68.0035 Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ___4____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Conduct of Operations Group # N/A K/A # G2.1.37 Importance Rating 4.3 Knowledge of procedures, guidelines, or limitations associated with reactivity management.

Question #67 Per ODP-ZZ-00001, Addendum 10, Reactivity Management / Operating Philosophy, a planned power change greater than a MINIMUM of ______ %

requires a reactivity management plan generated by reactor engineering.

A. 1 B. 3 C. 10 D. 15 Answer: C Explanation:

Per ODP-ZZ-00001 Addendum 10, Step 2.4.1 "When notified of a planned power transient greater than 10 %, Reactor Engineering will generate a Reactivity Plan based upon guidance from BEACON, SIMULATE and/or Xepred.xls based on the planned power reduction and rate."

A. Incorrect - plausible as there is an entire normal operating procedure addendum (OTG-ZZ-00004 Addendum 4) that outlines crew actions and responsibilities for power changes for less than 1%. Additionally, 1% is plausible as this is the conservative % (when compared to the other choices) and the general reactivity management philosophy is conservative in nature.

B. Incorrect - per step 2.1.6 "The CRS may also perform the RMSRO function when maneuvering the plant at a rate of < 3%/hr." Less than 3%/hr, the CRS can fulfill the role of the RMSRO function making 3% a plausible choice C. Correct D. Incorrect - Per step 2.1.4 "Decisions to maintain low power levels (less than 15%) for an extended period of time (greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ) are made utilizing APA-ZZ-01250, Operational Decision Making, to ensure Reactivity Management issues are considered." Additionally, power changes of >15% in one hour require chemistry sampling per TS SR requirements. Both of these reason make 15% a plausible choice.

Technical Reference(s):

1. ODP-ZZ-00001, Addendum 10, Reactivity Management / Operating Philosophy, Rev 26

2. OTG-ZZ-00004, Addendum 4, Less than 1% Power Changes from Full Power, Rev 3

NRC Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination: None Learning Objective: T61.003A, Normal Operations, LP#28, Objective A: EXPLAIN the following as applied in ODP ZZ 00001, Operations Dept. - Code of Conduct:

4. Reactivity management Question Source: Bank # ______

Modified Bank # ______

New ____X___

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Conduct of Operations Group # N/A K/A # G2.1.44 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 Per OTG-ZZ-00007, Refueling Preparation, Performance and Recovery, and ETP-ZZ-00035, Refueling Performance - IPTE, the reactor operator is required to ______________during core offload.

A. ensure Refuel Pool cleanup and Skimmer operation are secured B. maintain at least two source range neutron flux monitors operable C. maintain the locations of all fuel assemblies, fuel rods, and partial fuel rods D. ensure the Source Range Boron Dilution Flux Multiplication System (Flux Doubling) is reset Answer: B Explanation:

Per OTG-ZZ-00007 step 5.7.8 states " OFF-LOAD Fuel from Reactor to Spent Fuel Pool per ETP-ZZ-00035, Refueling Performance (IPTE)." making them both applicable procedures.

Per OTG-ZZ-00007 step 5.7.2 and 5.7.3 states "ESTABLISH Refuel Pool cleanup with demineralization per OTN-EC-00001 Add 4, Refuel Pool Cleanup Operations, and OTN-EC-00001 Add 5, Fuel Pool Cleanup Demineralize Operations. And 5.7.3. IF desired, ESTABLISH Refuel Pool Skimmer operation per OTN-EC-00001 Add 2, Spent Fuel Pool Skimmer Operation."

Per OTG-ZZ-00007 step 5.7.6 states " BLOCK boron dilution flux multiplication signals as follows to prevent spurious actuations during fuel movement:

a. PLACE the following switches in BLOCK:

PLACE SE HS-11, SR DOUBLED BLOCK/RESET in BLOCK.

PLACE SE HS-12, SR DOUBLED BLOCK/RESET in BLOCK.

And the Note prior to Step 5.7.6 states "The Source Range Boron Dilution Flux Multiplication System (Flux Doubling) is NOT required to be operable in MODE 6."

Per ETP-ZZ-00035, Refueling Performance, Section 4.0 4.1. Director, Nuclear Operations - Directs fuel handling operations, for both new and spent fuel, including the loading of the reactor core and all handling and storage of fuel on site.

NRC Written Examination Callaway Plant Reactor Operator 4.2. Shift Manager/Control Room Supervisor (SM/CRS) - Maintains plant conditions for core alterations.

4.3. Refueling Senior Reactor Operator (SRO) - Directs and Supervises core alterations and fuel movement.

4.4. Reactor Engineering Communicator - Directs crane operators for fuel movement, and maintains locations of all fuel assemblies, fuel rods, and partial fuel rods.

4.5. Reactor Operator (RO) - Monitors reactor for indications of criticality.

Per ETP-ZZ-00035, Refueling Performance, Section 5.5 During CORE ALTERATIONS, at least two source range neutron flux monitors are required to be OPERABLE per T/S LCO 3.9.3.

A. Incorrect - ODP-ZZ-00007 section 5.7 directs these be established if possible B. Correct - ETP-ZZ-00035, Refueling Performance, Section 5.5 During CORE ALTERATIONS, at least two source range neutron flux monitors are required to be OPERABLE per T/S LCO 3.9.3.

C. Incorrect - this is a responsibility of the Reactor Engineering Communicator as listed in ETP-ZZ-00035 D. Incorrect - ODP-ZZ-00007 section 5.7 directs these to be placed in the block position to prevent spurious actuation during fuel movement.

Technical Reference(s):

1. OTG-ZZ-00007, Refueling Preparation, Performance and Recovery, Rev 40

2. ETP-ZZ-00035, Refueling Performance - IPTE, Rev 45 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off-Normal Operations, LP#60, Objective B: STATE the Precautions and Limitations for OTG-ZZ-00007 Refueling Preparation, Performance, and Recovery.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3___

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Equipment Control Group # N/A K/A # G2.2.14 Importance Rating 3.9 Knowledge of the process for controlling equipment configuration or status.

Question #69 A component must be positioned out of its normal position in SUPPORT OF OPERATIONS.

It is expected that the component will be out of normal position past the end of shift.

Per ODP-ZZ-00035, Plant Status Control, .

1) When MUST the 50.59 evaluation be completed?

And

2) When is the LATEST that the Plant Status Control Tags be placed on the component?

A. 1) Before manipulating the component

2) Prior to the end of shift B. 1) Before manipulating the component

2) Prior to the component being manipulated C. 1) Within 60 days

2) Prior to the end of shift D. 1) Within 60 days

2) Prior to the component being manipulated Answer: A Explanation:

Per step 4.1.7, Plant Status Control Tags are generated in accordance with ODP-ZZ-00310, WPA and Caution Tagging. Tags may be placed at any time during the shift, but MUST be placed prior to the end of shift or immediately after turnover.

NRC Written Examination Callaway Plant Reactor Operator Per Step 4.2.1, REFER to Attachment 2 for flowchart for a visual depiction of when to apply the 50.59 Review process to PSC tagging. Per 4.2.3 IF the components are placed out-of-normal position in support of Maintenance or while waiting for WPA tags to be placed, GENERATE a Business Tracking CAR requiring a 10CFR50.59 evaluation to be performed within 60 days. Per 4.2.4, IF the components are placed out-of-normal position at the request of Operations personnel, COMPLETE a 10CFR50.59 evaluation immediately, before manipulating plant equipment.

Therefore the 50.59 evaluationmust be done before the component is manipulated but the tags must by the end of shift.

A. Correct B. Incorrect - the tags must be placed prior to the end of shift - before the component is manipulated is plausible as it applies to the 50.59 process.

C. Incorrect - the 50.59 must be done prior to the manipulation as this is in support of operations. If it was in support of maintenance it would be within 60 days which makes it plausible D. Incorrect - both are wrong Technical Reference(s):

1. ODP-ZZ-00035, Plant Status Control, Rev 17 References to be provided to applicants during examination: None Learning Objective: T61.003A, LP #8 Objective A: DESCRIBE the following as they pertain to ODP-ZZ-00035, Plant Status Control.

8. When 10CFR50.59 evaluations are required to be performed including time restrictions.

Question Source: Bank # __ ____

Modified Bank # ___X__L15416___

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Equipment Control Group # N/A K/A # G2.2.41 Importance Rating 3.5 Ability to obtain and interpret station electrical and mechanical drawings.

Question #70 Per station mechanical drawings, what picture below indicates a stop check valve?

A.

B.

C.

D.

Answer: B Explanation:

A. Incorrect - per M-220102 this is a needle valve B. Correct C. Incorrect - per M-220102 this is a packless globe valve D. Incorrect - per M-220102 this is a backflow preventer

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

1. M-220102, Symbols and Legend for System flow and P&IDs, Rev 9 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP - Print Reading Seminar, Enabling Objective #4:

UNDERSTAND the symbols associated with Mechanical Process and Instrumentation Diagrams Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____2____

10 CFR Part 55 Content:

10 CFR 55.41(b)(8)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Equipment Control Group # N/A K/A # G2.2.43 Importance Rating 3.0 Knowledge of the process used to track inoperable alarms.

Question #71 When defeating a main control board annunciator, a(n) ____(1)____ sticker is placed on the annunciator window. If the alarm is used in off normal or emergency operating procedures, issue a ____(2)____ describing redundant Control Room indications.

A. (1) orange (2) Night Order B. (1) orange (2) Standing Order C. (1) bright green (2) Night Order D. (1) bright green (2) Standing Order Answer: C Explanation:

Per section 4.5.1 of ODP-ZZ-00017, IF annunciator to be defeated is used in performance of off normal or emergency operating procedures, Refer To Appendix 1, Off Normal And Emergency Operation Procedure Related Annunciators and PREPARE a Night Order describing redundant Control Room indications.

Per 4.5.12 PLACE "Annunciator Defeat" (bright green) sticker on defeated annunciator window.

The Work Request sticker or Control Room Deficiency sticker is no longer required.

The orange sticker is a plausible distractor as per section 4.2 when a work request is submitted on an annunciator. Also a Standing Night order is used as a method of communication per ODP-ZZ-00008 A. Incorrect - wrong color B. Incorrect - both are wrong C. Correct D. Incorrect - wrong type of order

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

1. ODP-ZZ-00017, Annunciator Status and Tracking, Rev 37 References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems, LP #74, Objective E: DISCUSS the following as they pertain to ODP-ZZ-00017, "Annunciator Status and Tracking":

1. The purpose and scope

2. SM/OS responsibilities

3. Use of condition tags on annunciators associated with a Work Request

4. Defeat and restoration of:

a. Local annunciators

b. Main Control Board (MCB) annunciators

5. Tech Spec related annunciators IAW Appendix 2 Question Source: Bank # ___X L22896___

Modified Bank # ______

New _______

Question History: Last NRC Exam _______N/A_____

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____4____

10 CFR Part 55 Content:

10 CFR.55.41(b)(10)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Radiation Control Group # N/A K/A # G2.3.4 Importance Rating 3.2 Knowledge of radiation exposure limits under normal or emergency conditions.

Question #72 During a plant emergency, volunteers may receive one time dose in excess of their occupational Dose limit of up to _________ Rem for Category 1: Life Saving per HDP-ZZ-01450, Authorization to Exceed Federal Occupational Dose.

A. 10 B. 30 C. 50 D. 100 Answer: D Explanation:

Per HDP-ZZ-01450 Attachment 1, the dose limit for life savings activities is 100 rem DDE.

The distractors are also from Attachment 1 and Attachment 2. 10 rem and 30 rem are Category 2 limits for DDE and lens of the eye respectively. The 50 rem distractor is from attachment 2 and is the first listed dose that show health effects from a single acute Whole Body Absorbed Dose.

A. Incorrect - See above explanation.

B. Incorrect. - See above explanation.

C. Incorrect. - See above explanation.

D. Correct. - See above explanation.

Technical Reference(s):

1. HDP-ZZ-01450, Authorization to Exceed Federal Occupational Dose, Rev 13 References to be provided to applicants during examination: None.

Learning Objective: T61.003A, LP #33, Objective I: HDP-ZZ-01450, Authorization To Exceed Federal Occupational Dose

1. IDENTIFY who can authorize dose exposure in excess of 10CFR20.1201 dose limits.

NRC Written Examination Callaway Plant Reactor Operator

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 _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____2____

10 CFR Part 55 Content:

10 CFR 55.41(b)(12)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Radiation Control Group # N/A K/A # G2.3.15 Importance Rating 2.9 Knowledge to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Question #73 During a prejob briefing, Radiation Protection reports the following:

Electronic dosimeter (ED) dose alarm setting is 400 mrem

Electronic dosimeter (ED) dose rate alarm setting is 100 mrem/hr

Assigned RWP work area dose rate is 1000 mrem/hr Based on the conditions above, what describes your actions associated with the Radiological Control Area (RCA) and RWP?

A. Do not proceed due to an ED dose alarm B. Do not proceed due to an ED dose rate alarm C. Exit the RCA in 24 minutes due to an ED dose alarm D. Exit the RCA in 24 minutes due to an ED dose rate alarm Answer: B Explanation:

Based on the values given, the assigned dose rate is higher than the ED dose rate alarm setpoint, therefore the dose rate alarm would occur immediately when the area is entered. The distractors are a combination of the wrong alarm (EDs only have dose and dose rate alarms and 24 minutes which is 400/1000

60 minutes per hour).

A. Incorrect - wrong alarm type B. Correct C. Incorrect - both are wrong D. Incorrect - wrong time Technical Reference(s):

1. APA-ZZ-01004, Radiological Work standards, Rev 32

2. APA-ZZ-01000, Radiation protection program, Rev 46

NRC Written Examination Callaway Plant Reactor Operator References to be provided to applicants during examination: None Learning Objective: T61.003A, Normal Operations, LP#33 ALARA and RB Entry, Objective C:

APA-ZZ-01000, Callaway Plant Radiation Protection Program

1. DESCRIBE the purpose and scope

2. DESCRIBE the management policy regarding radiation exposure.

3. DEFINE:

a. Radiation Work Permit

b. Radiological Controlled Area (RCA)

c. Radiation Area

d. Specific Radiation Work Permit (SRWP)

e. General Radiation Work Permit (GRWP)

f. Occupational Radiation Exposure (ORE)

4. DISCUSS the Radiation Protection Stop Work Authority.

5. DISCUSS the Ameren Personnel exposure limits for the following categories:

a. Total Effective Dose Equivalent (TEDE)

b. Eye (LDE)

c. Total Organ Dose Equivalent (TODE)

d. Skin and Extremities (SDE)

6. EXPLAIN who can authorize requests for increased exposure.

7. DESCRIBE the actions to be taken in the event of a spill of radioactive material or other radiological incident.

8. DISCUSS actions when exceeding contamination limits.

Question Source: Bank # _X L16674_____

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2009_________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____2____

10 CFR Part 55 Content:

10 CFR 55.41(b)(12)

Comments:

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Emergency Procedures/Plan Group # N/A K/A # G2.4.3 Importance Rating 3.7 Ability to identify post-accident instrumentation.

Question #74 What instrument below is Post Accident Monitoring (PAM) Instrumentation?

A. B. C. D.

Answer: C Explanation:

Per OSP-SH-00001, Attachment 1 sheet 8 of 28, RCS Pressure - Wide Range is PAM instrumentation with 2 required channel of which BB PI-403 is one of the three total channels.

NRC Written Examination Callaway Plant Reactor Operator A. Incorrect - Emergency borate flow is not a required PAM instrument although plausible due to multiple boration actions in EOPs and boron ppm requirement of Technical Specifications.

B. Incorrect - PR NIs 41-44are not PAM instrumentation but PR NI 60/61 (Gammametrics) are which makes this a plausible distractor.

C. Correct - RCS Pressure - Wide Range is PAM Instrumentation per sheet 8 of OSP-SH-00001 D. Incorrect - Containment recirc level is not a PAM instrument but plausible as the CTMT Normal sump level (meter adjacent to Recirc sump Level) is a PAM Instrument. Additionally, the RWST Instruments are also PAM instrumentation and as the Recirc sump level and RWST levels measure water inventory during EOP use, it may be believed Recirc Sump level is PAM Instrumentation.

Technical Reference(s):

1. OSP-SH-00001, PAM Channel Check, Rev 30 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP #30 Objective I: PARAPHRASE Technical Specification requirements for Reactor Instrumentation from TS Section 3.3 Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ___3____

10 CFR Part 55 Content:

Comments:

Callaway does not have a specific method (Dot on label plate, hash marks around instrument, etc) to identify PAM instrumentation on the MCBs.

NRC Written Examination Callaway Plant Reactor Operator Examination Outline Cross-reference: Level RO Rev 0 Tier # 3 Emergency Procedures/Plan Group # N/A K/A # G2.4.17 Importance Rating 3.9 Knowledge of EOP terms and definitions.

Question #75 Per ODP-ZZ-00025, EOP/OTO User's Guide, what is the definition of "TRY" when used in EOP steps?

A. to make a continued effort when success may not be immediately obtainable B. to turn on or turn off, as necessary, to achieve the stated objective or function C. to continuously control a given plant parameter within a specified range or other specified limit as required by the instruction D. to determine if a condition exists and if it does not, this word denotes permission and direction to take the steps necessary to make the condition exist Answer: A Explanation:

Per Attachment 1 of ODP-ZZ-00025, Try is defined as "to make a continued effort when success may not be immediately obtainable."

All of the distractors are action verbs and defined in the same ODP-ZZ-00025 Attachment 1.

A. Correct B. Incorrect - this is the definition of Operate C. Incorrect - this is the definition of Maintain D. Incorrect - this is the definition of Ensure Technical Reference(s):

1. ODP-ZZ-00025, EOP/OTO User's Guide, Rev 37 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP#1, Objective AA: DESCRIBE the General Procedural Guidance provided by ODP-ZZ-00025, EOP/OTO Users Guide.

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

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge ___X__

Comprehension or Analysis _____

LOD _____3_____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 Small Break LOCA Group # 1 K/A # 00009 G2.4.30 Importance Rating 4.1 Knowledge of events related to system operation / status that must be reported to internal organizations or external agencies, such as the State, the NRC, or the transmission system operator.

Question #76 Reactor power was 100% when a small break Loss of Coolant Accident (LOCA) occurred resulting in a reactor trip and Safety Injection (SI).

Excluding EAL notifications and per APA-ZZ-00520, Reporting Requirements and Responsibilities, the above event is required to be reported as __(1)___ to the NRC within a MAXIMUM time of ___(2)___?

A. (1) Valid system actuation (2) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months B. (1) Valid system actuation (2) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C. (1) RPS actuation while the reactor is critical (2) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months D. (1) RPS actuation while the reactor is critical (2) 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months Answer: C Explanation:

Per APA-ZZ-00520, a reactor trip and safety injection is a 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months report because "Immediate Notification - RPS actuation while the reactor is critical - 10CFR50.72(b)(2)(iv)(B)". 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is plausible for several reasons which include:

Immediate Notification - Valid system actuation - 10CFR50.72(b)(3)(iv)(A)

Immediate Notification - Nuclear plant, including principal safety barriers, seriously degraded - 10CFR50.72(b)(3)(ii)(A)

A. Incorrect - See above explanation B. Incorrect - See above explanation C. Correct D. Incorrect - See above explanation

NRC Written Examination Callaway Plant Senior Reactor Operator Technical Reference(s):

1. APA-ZZ-00520, Reporting Requirements and Responsibilities, Rev 53

2. ODP-ZZ-00001, Addendum 13, Shift Manager Communications, Rev 24 References to be provided to applicants during examination: None Learning Objective: T61.003A, Normal Operations, LP#31, Objective #B, PERFORM the following as they pertain to APA-ZZ-00520, REPORTING REQUIREMENTS AND RESPONSIBILITIES:

1. DESCRIBE the purpose and scope

2. DISCUSS the incidents reportable in the following time frames:

a. 15 minutes

b. 30 minutes

c. Immediate (1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months )

d. Immediate (2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months )

e. Immediate (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months )

f. Immediate (8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months )

g. 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months notification

h. 48 hour5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months notification

i. 5 day

j. As soon as Possible/Promptly

3. DESCRIBE the 10 CFR 50.72 Immediate Notification Requirements Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____4_____

10 CFR Part 55 Content:

Comments:

SRO ONLY due to 10 CFR 55.43(b)(1) - Conditions and limitations in the facility license. SROs are required to know and fulfill this part of the facility license.

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 Loss of Offsite Power Group # 1 K/A # 00056 G2.2.25 Importance Rating 4.2 Knowledge of the bases in Technical Specifications for limiting conditions for operation and safety limits Question #77 Per 3.8.1 Technical Specifications Basis, when BOTH Offsite Circuits are inoperable, why is a completion time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months to restore one offsite circuit to operable status allowed?

A. all redundancy in the AC electrical power supplies has been lost B. there is insufficient AC sources available to power the minimum required ESF functions C to allow time to evaluate and repair any inoperability of required redundant safety features D. the configuration of the redundant AC electrical power systems that remain available is NOT susceptible to a single failure Answer: D Explanation:

A. Incorrect - This is the basis of 3.8.1 Condition H for the condition when 3 or more AC sources are inoperable. Page 3.8.1 -14 B. Incorrect - This is the basis of 3.8.1 Condition E for the condition when 2 EDGS are inoperable. Page 3.8.1 -13 C. Incorrect - This is the basis of the completion time and required action in 3.8.1 Condition A.2 and C.1 for the "required redundant features that are inoperable" section. Its basis states "to allow the operator time to evaluate and repair any discovered inoperabilities" The wording was paraphrased to make it creditable. While this action appears in C.1 (BOTH offsite circuits inoperable), this is action C.1 with a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from discovery completion time which is not what the question stem is asking (i.e. is asking about the 24 hr clock) therefore making it wrong.

D. Correct - This is the basis of 3.8.1 Condition C for the condition when 2 Offsite circuits are inoperable (i.e Loss of offsite power). Page B 3.8.1-11.

Technical Reference(s):

1. Technical Specification 3.8.1 and it bases

NRC Written Examination Callaway Plant Senior Reactor Operator References to be provided to applicants during examination: None Learning Objective: T61.0110 Systems, LP #1 Objective G: DESCRIBE the two offsite independent circuits which satisfy the Tech Specs offsite power requirements.

Question Source: Bank # __X____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____2019 Q#79________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

Comments:

k/a match as the question asks the basis for BOTH offsite circuit operable but written in the positive voice. The LCO 3.8.1 C for BOTH offsite circuit inoperable is in effect the "loss of offsite power" and the question examines the TS basis of LCO 3.8.1 Condition C.

SRO ONLY due to the knowledge of Facility operating limitations in the technical specifications and their bases.

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 Loss of DC Power Group # 1 K/A # 000058 AA2.03 Importance Rating 3.9 Ability to determine and interpret the following as they apply to the Loss of DC Power: DC loads lost; impact on ability to operate and monitor plant systems Question #78 Reactor Power is 6%.

'A' Emergency Diesel Generator (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) How should the crew secure the NE01, 'A' Emergency Diesel Generator (EDG)?

And (2) What procedure should be implemented by the CRS?

A. (1) Locally OPEN NB0111, NB01 EMERG FEED FROM A STBY DG NE01, and then CLOSE the 'A' EDG fuel racks (2) OTO-NK-00001, Failure of NK Battery Charger, Attachment A, Loss of Power To NK01 B. (1) Locally OPEN NB0111, NB01 EMERG FEED FROM A STBY DG NE01, and then CLOSE the 'A' EDG fuel racks (2) OTO-NK-00002, Loss of Vital 125 VDC Bus, Attachment A, Loss of Power To NK01 C. (1) PLACE KJ HS-9, NE01 Local Master Transfer Switch, To LOCAL/MAN Position and DEPRESS the Emergency Stop PB, KJ HS-8B (2) OTO-NK-00001, Failure of NK Battery Charger, Attachment A, Loss of Power To NK01 D. (1) PLACE KJ HS-9, NE01 Local Master Transfer Switch, To LOCAL/MAN Position and DEPRESS the Emergency Stop PB, KJ HS-8B

NRC Written Examination Callaway Plant Senior Reactor Operator (2) OTO-NK-00002, Loss of Vital 125 VDC Bus, Attachment A, Loss of Power To NK01 Answer: B Explanation:

The correct method of securing the EDG without control power is per OTO-NK-00002 Attachment A step #1 RNO which states:" Locally STOP EDG A by performing the following:

a) OPEN NB01 EMERG FEED FROM A STBY DG NE01: NB0111 b) CLOSE the EDG A fuel racks until diesel generator stops.

PLACE NE01 Local Master Transfer Switch To LOCAL/MAN Position (KJ HS-9) does several functions and is Attachment A step #2. This action is in several procedures for several reasons but basically transfers control to the local panel (assuming there is control power) and inhibits autostart functions. To make it a plausible choice, the EDG the local emergency stop PB must be depressed also. This series of actions also appears in ECA-0.0 step #5 RNO as follows:

a) Locally shutdown associated Diesel Generator as follows:

1. Place Master Transfer Switch to LOC/MAN:: KJ HS-9

2. STOP the Diesel: KJ HS-8B Making it a plausible choice but incorrect as there is no control power available and these actions would not secure the EDG.

A. Incorrect - correct action but incorrect procedure choice however plausible since annunciator 25C is an entry condition into OTO-NK-00001 B. Correct annunciators 25B & C are entry conditions into OTO-NK-00002 and the correct method of securing the EDG as listed above.

C. Incorrect - both are wrong D. Incorrect - correct procedure choice but incorrect action to secure the EDG as listed above.

Technical Reference(s):

1. OTO-NK-00002, Loss of Vital 125 VDC Bus, Rev 016 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP#50, Objective C: Given a set of plant conditions or parameters indicating a Loss of Vital 125 VDC Bus, IDENTIFY the correct procedure(s) to be utilized and OUTLINE the high level actions to stabilize the plant.

Question Source: Bank # _______

Modified Bank # _X____L16148___

New _______

Question History: Last NRC Exam ___N/A______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

NRC Written Examination Callaway Plant Senior Reactor Operator Comprehension or Analysis __X__

LOD ____3______

10 CFR Part 55 Content:

Comments:

SRO ONLY due to ES401 Figure 2 of NUREG 1021 as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely by knowing immediate operator actions? NO Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? NO Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures-

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 Generator Voltage and Electric Grid Group # 1 Disturbances K/A # 000077 AA2.05 Importance Rating 3.8 Ability to determine and interpret the following as they apply to l Generator Voltage and Electric Grid Disturbances: Operational status of offsite circuit REFERENCE PROVIDED Question #79 Reactor power is 100%.

The A EDG, NE01, was tagged out for lube oil system maintenance 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months ago.

An Electrical Grid Disturbance is in progress.

The Transmission Operations Supervisor reports that a Category 8 Alarm is received. The Predicted Contingency Voltage is 325 kV.

In order to remain in compliance with the Technical Specifications, the reactor must be in MODE 3 within a MAXIMUM of _______ hours?

A. 7 B. 14 C. 18 D. 30 Answer: A Explanation Per Attachment 3 of OSP-NE-00003, If the Predicted Voltage is outside of the required voltage range, the SM/CRS shall declare the offsite circuits INOPERABLE. Per of OSP-NE-00003, the Contingency Analysis Computer Calculated Operability Limit is in the 372.6 - 329.8 kV. Therefore with a predicted analysis point less than 329.8 kV, the SRO candidate should declare both offsite circuits inop.

Based on the given plant conditions, Tech Spec 3.8.1 Conditions B, C, D, H are not met. The shortest time duration would be for 3.8.1. H which directs you to enter T.S. 3.0.3 immediately which requires Mode 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months .

NRC Written Examination Callaway Plant Senior Reactor Operator 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months is plausible if the candidate does not process that both offsite circuits are inop and proceeds with only Condition B (one DG inop) not met and then progresses to 3.8.1 G to be in mode 3 within an additional 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . There are 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months left on the original time clock for condition B, therefore 24 + 6 = 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . The candidate could also arrive at the 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months distractor by only applying 3.8.1C which has a 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months completion time. This time added with the 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of 3.8.1 G would also yield a 30 hour3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months time limit.

If the candidate applies 3.8.1 D (1 DG and 1 offsite circuit inop) as limiting (incorrectly), there is a 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months completion time before 3.8.1 G is applied. This would give the candidate a calculated time of 18 hours2.083333e-4 days 0.005 hours 2.97619e-5 weeks 6.849e-6 months . (12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of D + 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months of G).

If the candidate applies the note in 3.8.1 condition D to enter LCO 3.8.9 condition A to restore within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months and then proceeds to applies 3.8.9 Condition D to be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , the calculated time would be 14 hours1.62037e-4 days 0.00389 hours 2.314815e-5 weeks 5.327e-6 months . While this is a correct application it is not the most limiting time to be in Mode 3.

A. Correct B. Incorrect - not the most limiting time applied 3.8.9 A then D C. Incorrect - not the most limiting time applied 3.8.1 D then G D. Incorrect - not the most limiting time applied 3.8.1 B then G (or C then G)

Technical Reference(s):

1. OSP-NE-00003, Technical Specifications Actions - A.C. Sources Rev 34

2. Technical Specification Section 3.8, Amendment #202 References to be provided to applicants during examination:

1. Technical Specification Section 3.8.1, AC Sources Operating

2. Technical Specification Section 3.8.9, Distribution Systems Learning Objective: T61.0110 Systems, LP #6, Objective G: EXPLAIN the Technical Specifications and bases for the Safeguards Power System Question Source: Bank # _X___No ID_____

Modified Bank # ______

New _______

Question History: Last NRC Exam __2014_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

LOD _____2_____

NRC Written Examination Callaway Plant Senior Reactor Operator 10 CFR Part 55 Content:

10 CFR 55.55.43(b)(2)

Comments:

SRO Only due to 43(b) #2 - Facility operating limitations in the technical specifications and their bases.

Additionally per Figure 1 Attachment 2 of ES-401,

Can question be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS/TRM Action? Is No

Can question be answered solely by knowing the LCO/TRM information listed above-the-line? is No

Can question be answered solely by knowing the TS Safety Limits? Is No

Does the question involve one or more of the following for TS, TRM, or ODCM?

Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1) is Yes which means this is an SRO Only question

AC Sources - Operating 3.8.1 3.8 ELECTRICAL POWER SYSTEMS 3.8.1 AC Sources - Operating LCO 3.8.1 The following AC electrical sources shall be OPERABLE:

a. Two qualified circuits between the offsite transmission network and the onsite Class 1E AC Electrical Power Distribution System; and

b. Two diesel generators (DGs) capable of supplying the onsite Class 1E power distribution subsystem(s); and

c. Load Shedder and Emergency Load Sequencer (LSELS) for Train A and Train B.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS

NOTE -----------------------------------------------------------

LCO 3.0.4.b is not applicable to DGs.

COMPLETION CONDITION REQUIRED ACTION TIME A. One offsite circuit A.1 Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. OPERABLE offsite circuit. AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND A.2 ------------ NOTE -----------

In MODES 1, 2, and 3, the turbine driven auxiliary feedwater pump is considered a required redundant feature.

(continued)

CALLAWAY PLANT 3.8-1 Amendment No. 199

AC Sources - Operating 3.8.1 ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One offsite circuit Declare required 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months from inoperable. feature(s) with no offsite discovery of no (continued) power available offsite power to one inoperable when its train concurrent redundant required with inoperability of feature(s) is inoperable. redundant required feature(s)

AND A.3 Restore offsite circuit to 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months OPERABLE status.

AND 6 days from discovery of failure to meet LCO B. One DG inoperable. B.1 Perform SR 3.8.1.1 for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the offsite circuit(s).

AND Once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months thereafter AND B.2 ------------ NOTE -----------

In MODES 1, 2, and 3, the turbine driven auxiliary feedwater pump is considered a required redundant feature.

(continued)

CALLAWAY PLANT 3.8-2 Amendment 133

AC Sources - Operating 3.8.1 ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME B. One DG inoperable. Declare required 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months from (continued) feature(s) supported by discovery of the inoperable DG Condition B inoperable when its concurrent with required redundant inoperability of feature(s) is inoperable. redundant required feature(s)

AND B.3.1 Determine OPERABLE 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months DG is not inoperable due to common cause failure.

OR B.3.2 ------------ NOTE -----------

The required ACTION of B.3.2 is satisfied by the automatic start and sequence loading of the diesel generator.

Perform SR 3.8.1.2 for 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months OPERABLE DG.

AND (continued)

CALLAWAY PLANT 3.8-3 Amendment 186

AC Sources - Operating 3.8.1 ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME B. One DG inoperable. B.4 Restore DG to ---------NOTE---------

(continued) OPERABLE status. A one-time Completion Time of 14 days is allowed to support planned replacement of ESW B train piping prior to April 30, 2009.

72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND 6 days from discovery of failure to meet LCO C. Two offsite circuits C.1 ------------ NOTE -----------

inoperable. In MODES 1, 2, and 3, the turbine driven auxiliary feedwater pump is considered a required redundant feature.

Declare required 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months from feature(s) inoperable discovery of when its redundant Condition C required feature(s) is concurrent with inoperable. inoperability of redundant required features AND (continued)

CALLAWAY PLANT 3.8-4 Amendment 199

AC Sources - Operating 3.8.1 ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME C. Two offsite circuits C.2 Restore one offsite circuit 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months inoperable. to OPERABLE status.

(continued)

D. One offsite circuit -------------------- NOTE -------------------

inoperable. Enter applicable Conditions and Required Actions of LCO 3.8.9, AND Distribution Systems - Operating, when Condition D is entered with no One DG inoperable. AC power source to any train.

D.1 Restore offsite circuit to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OPERABLE status.

OR D.2 Restore DG to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months OPERABLE status.

E. Two DGs inoperable. E.1 Restore one DG to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months OPERABLE status.

F. One required LSELS F.1 Declare the affected DG Immediately inoperable. and offsite circuit inoperable.

AND F.2 Restore required LSELS 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to OPERABLE status.

(continued)

CALLAWAY PLANT 3.8-5 Amendment 186 l

AC Sources - Operating 3.8.1 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME G. Required Action and G.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition A, B, C, D, E, AND or F not met.

G.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months H. Three or more AC sources H.1 Enter LCO 3.0.3. Immediately inoperable.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.1.1 Verify correct breaker alignment and indicated power In accordance availability for each required offsite circuit. with the Surveillance Frequency Control Program (continued)

CALLAWAY PLANT 3.8-6 Amendment No. 202

Distribution Systems - Operating 3.8.9 3.8 ELECTRICAL POWER SYSTEMS 3.8.9 Distribution Systems - Operating LCO 3.8.9 Train A and Train B AC, DC, and AC vital bus electrical power distribution subsystems shall be OPERABLE.

APPLICABILITY: MODES 1, 2, 3, and 4.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One AC electrical power A.1 Restore AC electrical 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months distribution subsystem power distribution inoperable. subsystem to AND OPERABLE status.

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO B. One AC vital bus B.1 Restore AC vital bus 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months subsystem inoperable. subsystem to OPERABLE status. AND 16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO (continued)

CALLAWAY PLANT 3.8-38 Amendment No. 202 l

Distribution Systems - Operating 3.8.9 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME C. One DC electrical power C.1 Restore DC electrical 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months distribution subsystem power distribution inoperable. subsystem to AND OPERABLE status.

16 hours1.851852e-4 days 0.00444 hours 2.645503e-5 weeks 6.088e-6 months from discovery of failure to meet LCO D. Required Action and D.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time not met. AND D.2 Be in MODE 5. 36 hours4.166667e-4 days 0.01 hours 5.952381e-5 weeks 1.3698e-5 months E. Two trains with inoperable E.1 Enter LCO 3.0.3. Immediately distribution subsystems that result in a loss of safety function.

SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.8.9.1 Verify correct breaker alignments and voltage to In accordance required AC, DC, and AC vital bus electrical power with the distribution subsystems. Surveillance Frequency Control Program CALLAWAY PLANT 3.8-39 Amendment No. 202

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 Loss of Nuclear Service Water Group # 1 K/A # 000062 G2.1.32 Importance Rating 4.0 Ability to explain and apply system limits and precautions Question #80 The plant is in a winter maintenance outage with the following conditions:

Service Water pumps A and B are RUNNING

Essential Service Water pumps are SECURED since the UHS pond has been pumped down Service water pump discharge pressure is approximately 85 psig, the CRS should direct what section of OTN EA-00001, Service Water System, and why?

A. Section 5.9, Operation During Extreme Cold Weather To ensure discharge pressure does not exceed 85 psig for a MAXIMUM of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months B. Section 5.9, Operation During Extreme Cold Weather To ensure discharge pressure does not exceed 85 psig for a MAXIMUM of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months C. Section 5.4, Securing a Service Water Pump To ensure discharge pressure does not exceed 85 psig for a MAXIMUM of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months D. Section 5.4, Securing a Service Water Pump To ensure discharge pressure does not exceed 85 psig for a MAXIMUM of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Answer: C Explanation:

As there is no emergency in progress (hence EOP and EOP addendums don't apply) and the plant doesn't not have an offnormal procedure for the service water system, OTN-EA-00001 (the normal operating procedure) is the appropriate procedure to address the situation. The purpose of OTN-EA-00001 is to provide instructions for proper alignment and startup of the Service Water System and its scope applies to the Service Water System and its interconnections.

OTN-EA-00001 states:

3.1.4. During periods of normal system operation, the discharge pressure of any pump should NOT exceed 80 psig. A pump should be stopped if the discharge pressure exceeds 80 psig.

3.1.5. During winter months when the UHS pond is pumped down, Service Water Pump discharge pressure can exceed 80 psig, due to most of the Temperature Control Valves being closed. Short term operation of the system above 80 psig, up to a maximum of 85 psig, is acceptable for up to 2-hours within a 24-hour period.

NRC Written Examination Callaway Plant Senior Reactor Operator A. Incorrect but plausible since this section deals with keeping system pressure below 80 psig by changing CCW HX alignment. The time aspect is correct per step 3.1.5 (above).

B. Incorrect plausible since this section deals with keeping system pressure below 80 psig by changing CCW HX alignment and time is wrong. 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is plausible as step 3.1.5 discusses a 24 rolling period.

C. Correct since correct procedure and section and right time D. Incorrect plausible since correct procedure section but the wrong time as discussed above.

Technical Reference(s):

1. OTN-EA-00001, Service Water System, Rev 041 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#4, Objective H: Describe Circ. and Service Water Pump operations

1. Startup

2. Shutdown

3. Discharge Valve operation Question Source: Bank # ______

Modified Bank # ______

New ____X___

Question History: Last NRC Exam _____NA_____

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD _____2_____

10 CFR Part 55 Content:

Comments:

SRO ONLY due to ES401 Figure 2 of NUREG 1021 as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely by knowing immediate operator actions? NO

NRC Written Examination Callaway Plant Senior Reactor Operator Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? NO Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures-

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 Loss of Emergency Coolant Recirc. / 4 Group # 1 K/A # W/E11 EA2.1 Importance Rating 4.2 Ability to determine and interpret the following as they apply to the (Loss of Emergency Coolant Recirculation): Facility conditions and selection of appropriate procedures during abnormal and emergency operations.

Question #81 The control room crew has just transitioned to ECA-1.1, Loss of Emergency Coolant Recirculation from ECA-1.2, LOCA Outside Containment.

Current plant conditions are as follows:

NB01 is deenergized

RCS Pressure is 1500 psig slowly lowering

RCS Subcooling is 60°F

RWST level is 10% stable

'A' and 'D' RCPs are in service Based on the above conditions, what is the NEXT procedure the crew would utilize?

A. EOP Addendum 3, Starting an RCP B. OTN-GS-00001, Containment hydrogen Control System C. EOP Addendum 16, Placing Hydrogen Analyzers In Service D. OTN-BG-00002, Reactor Makeup Control and Boron Thermal Regeneration System Answer: D Explanation:

While RCS subcooling supports starting an RCP (but not terminating SI which adds to the plausibility of a different distractor), there is no need to start an additional RCP, The combination of A and D RCP will allow normal PZR spray to be utilized to depressurize the RCS. Plausible as the candidate may believe the procedural requirement is to start all available RCPs to ensure core heat removal as ECCS availability is limited by RWST level.

OTN-GS-00001 is plausible since H2 concentration would rise during a LOCA EOP Addendum 16 is plausible since the conditions provided will result in loss of core decay heat removal from ECCS injection flow. Step 39 places the analyzer into service for monitoring fuel

NRC Written Examination Callaway Plant Senior Reactor Operator conditions however actions at step 31 (further directions to implement OTN-BG-00002) would precede step 39 for mitigation strategy therefore EOP ADD 16 is not the NEXT priority.

A. Incorrect - See above explanation B. Incorrect - See above explanation C. Incorrect - See above explanation D. Correct - as RWST level is low, the next priority is to makeup to RWST as necessary per ECA-1.1 step #9.

Technical Reference(s):

1. ECA-1.1, Loss of Emergency Coolant Recirculation, Rev 14

2. OTN-BG-00002, Reactor Makeup Control and Boron Thermal Regeneration System, Rev 51 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP #13, Objective 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 ___2017 NRC Question #80________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

LOD ____3____

10 CFR Part 55 Content:

Comments:

SRO ONLY due to ES401 Figure 2 of NUREG 1021 as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely by knowing immediate operator actions? NO Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? NO

NRC Written Examination Callaway Plant Senior Reactor Operator Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures YES

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 Inoperable Stuck Control Rod Group # 1 K/A # 000005 G2.2.40 Importance Rating 4.7 Ability to apply Technical Specifications for a system REFERENCE PROVIDED Question #82 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 Which of the below satisfies the required action(s) of TS LCO 3.1.4?

A. Restore rods to within alignment limits within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months or be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months B. Verify Shutdown Margin to be within the limits provided in the COLR within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and be in Mode 3 within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months C. Initiate action within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the unit in Mode 3 within 7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months and Mode 4 within 13 hours1.50463e-4 days 0.00361 hours 2.149471e-5 weeks 4.9465e-6 months and Mode 5 within 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months D. Verify Shutdown Margin to be within the limits provided in the COLR within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months , reduce thermal power to < 75% RTP within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , verify Shutdown Margin to be within the limits provided in the COLR once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months , perform SRs 3.2.1.1, 3.2.1.2, and 3.2.2.1 within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months , and re-evaluate safety analyses and confirm results remain valid for duration of operation under these conditions within 5 days Answer: B Explanation: Note: there are 2 misaligned rods in the stem data: D2 and B12

NRC Written Examination Callaway Plant Senior Reactor Operator A. Incorrect. Plausible as misalignment is allowed for 1 rod; not allowed for more than 1 rod. This distractor is a combination of TS LCO actions B.1 and C.1.

B. Correct. Correct action to take IAW TS 3.1.4 Action D for more than one rod not within alignment limits. This answer is Required Action D.1.1 and D.2.

C. Incorrect. Plausible if it is believed that required actions of 3.1.4 do not "cover" the situation described in the stem and it is believed that LCO 3.0.3 should be applied. The distractor is the entire statement of LCO actions a. b. and c.

D. Incorrect. Plausible as misalignment is allowed for 1 rod; but not allowed for more than 1 rod and as the LCO does not state a "separate entry condition is allowed for each.." which makes it incorrect. This distractor statement is LCO 3.1.4 required actions B.2.1.1, B.2.2, B.2.3, B2.4.4, B.2.4.5 and B.2.4.6 Technical Reference(s):

1. Technical Specifications 3.1, Reactivity Control Systems,

2. Technical Specifications 3.1.4 Rod Group Alignment Limits References to be provided to applicants during examination:

1. Technical Specifications 3.1.4 Rod Group Alignment Limits Learning Objective: T61.0110, System, LP-26, Objective U, State the Technical Specification limiting conditions for operations (LCOs) applicable to the rod control system.

Question Source: Bank # _X___No ID____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____2013_____

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3______

10 CFR Part 55 Content:

NRC Written Examination Callaway Plant Senior Reactor Operator Comments:

SRO Only due to 43(b) #2 - Facility operating limitations in the technical specifications and their bases.

Additionally per Figure 1 Attachment 2 of ES-401,

Can question be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS/TRM Action? Is No

Can question be answered solely by knowing the LCO/TRM information listed above-the-line? is No

Can question be answered solely by knowing the TS Safety Limits? Is No

Does the question involve one or more of the following for TS, TRM, or ODCM?

Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1) knowledge of TS bases that is required to analyze TS-required actions and terminology is Yes which means this is an SRO Only question

Rod Group Alignment Limits 3.1.4 3.1 REACTIVITY CONTROL SYSTEMS 3.1.4 Rod Group Alignment Limits LCO 3.1.4 All shutdown and control rods shall be OPERABLE.

AND Individual indicated rod positions shall be within 12 steps of their group step counter demand position.

APPLICABILITY: MODES 1 and 2.

ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME A. One or more rod(s) A.1.1 Verify SDM to be within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months inoperable. the limits provided in the COLR.

OR A.1.2 Initiate boration to restore 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SDM to within limit.

AND A.2 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months (continued)

CALLAWAY PLANT 3.1-7 Amendment No. 133

Rod Group Alignment Limits 3.1.4 ACTIONS (continued)

COMPLETION CONDITION REQUIRED ACTION TIME B. One rod not within B.1 Restore rod to within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months alignment limits. alignment limits.

OR B.2.1.1 Verify SDM to be within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months the limits provided in the COLR.

OR B.2.1.2 Initiate boration to restore 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months SDM to within limit.

AND B.2.2 Reduce THERMAL 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months POWER to 75% RTP.

AND B.2.3 Verify SDM to be within Once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months the limits provided in the COLR.

AND B.2.4 Perform SR 3.2.1.1 and 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months SR 3.2.1.2.

AND B.2.5 Perform SR 3.2.2.1. 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months AND (continued)

CALLAWAY PLANT 3.1-8 Amendment No. 133

Rod Group Alignment Limits 3.1.4 ACTIONS COMPLETION CONDITION REQUIRED ACTION TIME B. (continued) B.2.6 Re-evaluate safety 5 days analyses and confirm results remain valid for duration of operation under these conditions.

C. Required Action and C.1 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months associated Completion Time of Condition B not met.

D. More than one rod not D.1.1 Verify SDM to be within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months within alignment limit. the limits provided in the COLR.

OR D.1.2 Initiate boration to restore 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months required SDM to within limit.

AND D.2 Be in MODE 3. 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months CALLAWAY PLANT 3.1-9 Amendment 133

Rod Group Alignment Limits 3.1.4 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.1.4.1 Verify individual rod positions within alignment limit. In accordance with the Surveillance Frequency Control Program SR 3.1.4.2 Verify rod freedom of movement (trippability) by In accordance moving each rod not fully inserted in the core with the 10 steps in either direction. Surveillance Frequency Control Program SR 3.1.4.3 Verify rod drop time of each rod, from the fully Prior to reactor withdrawn position, is 2.7 seconds from the criticality after beginning of decay of stationary gripper coil voltage each removal of to dashpot entry, with: the reactor head

a. Tavg 500°F; and

b. All reactor coolant pumps operating.

CALLAWAY PLANT 3.1-10 Amendment No. 202

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 Accidental Liquid Radwaste Release Group # 2 K/A # 000059 G2.2.25 Importance Rating 4.2 Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits.

Question #83 An unisolable leak of the Refueling Water Storage Tank (RWST) has occurred.

What is the RWST FSAR radioactive materials limit and the basis for that limit?

A. 50 Curies to assure that an uncontrolled release of the tanks contents would be less than the limits of 10CFR20.

B. 50 Curies to assure that the radiation levels in the accessible areas surrounding the tank are less than 100 MRem / hour.

C. 150 Curies to assure that an uncontrolled release of the tanks contents would be less than the limits of 10CFR20.

D. 150 Curies to assure that the radiation levels in the accessible areas surrounding the tank are less than 100 MRem / hour.

Answer: C Explanation:

FSAR 16.11.1.5, The quantity of radioactive material contained in each of the following unprotected outdoor tanks shall be limited to less than or equal to 150 Curies, excluding tritium and dissolved or entrained noble gases for Reactor Makeup Water Storage Tank Restricting the quantity of radioactive material contained in the RWST provides assurance that in the event of an uncontrolled release of the tanks' contents, the resulting concentrations would be less than the limits of 10 CFR Part 20 A. Incorrect, limit is 150 curies B. Incorrect 100 mrem/hr is annual TEDE dose to the public for gaseous effluents C. Correct see above D. Incorrect 100 mrem/hr is annual TEDE dose to the public for gaseous effluents Technical Reference(s):

1. FSAR 16.11.1.5, Liquid Holdup Tanks Limiting Condition for Operation

2. APA-ZZ-01003, ODCM, Rev 024

NRC Written Examination Callaway Plant Senior Reactor Operator References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#13, Objective E: STATE the LCOs, bases and surveillance operability requirements associated with RWST Technical Specifications.

Question Source: Bank # __L14601_

Modified Bank # ______

New _______

Question History: Last NRC Exam _____N/A_______

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD _____3_____

10 CFR Part 55 Content:

Comments:

SRO Only due to 43(b) #2 - Facility operating limitations in the technical specifications and their bases.

Additionally per Figure 2-1 Attachment 2 of ES-401,

Can question be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS/TRM Action? Is No

Can question be answered solely by knowing the LCO/TRM information listed above-the-line? is No

Can question be answered solely by knowing the TS Safety Limits? Is No

Does the question involve one or more of the following for TS, TRM, or ODCM?

Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1) knowledge of TS bases that is required to analyze TS-required actions and terminology is Yes which means this is an SRO Only question

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 Plant Fire on site Group # 2 K/A # 000067 AA2.13 Importance Rating 4.4 Ability to determine and interpret the following as they apply to the Plant Fire on Site: Need for emergency plant shutdown Question #84 Reactor Power is 100%.

KC008 alarms. A control room fire alarm is present.

At RL019, the 'C' and 'D' CCW pumps are spuriously starting and stopping.

Control Boards are hard to read due to smoke/haze forming in the control room.

Based on these conditions, the Control Room Supervisor should direct a control room operator to perform ?

A. OTO-ZZ-00002, Control Room Operations with Fire, Attachment A, Control Room Panels B. OTO-ZZ-00002, Control Room Operations with Fire, Attachment B, Pre-Evacuation Contingency Actions C. OTO-ZZ-00001, Control Room Inaccessibility, Attachment B, Balance Of Plant (BOP) Operator Actions With Fire D. OTO-ZZ-00001, Control Room Inaccessibility, Attachment F, Manual Start Of DG NE02 And Energizing NB02 Answer: C Explanation:

Beginning at OTO-KC-00001 Addendum C-27 step 1.3 "If Spurious Operation of equipment is observed, consider going to OTO-ZZ-00001, Control Room Inaccessibility." However, step #1.4 states "For a fire in the Control Room refer to OTO-ZZ-00002, Control Room Operations with Fire". If the CRS decides to implement OTO-ZZ-00002, at step #2 the RNO applies which states "IF determined by SM/CRS that Control Room evacuation is required, THEN Go To OTO-ZZ-00001, Control Room Inaccessibility, Step 1."

To summarize, with the given data in the stem, OTO-ZZ-00001 implementation is the correct procedure path.

NRC Written Examination Callaway Plant Senior Reactor Operator A. Incorrect - OTO-ZZ-00002, Attachment A, Control Room Panels, is plausible but incorrect as this is directed in step 6.b of OTO-ZZ-00002 which states "Refer To Attachment A, Control Room Panels, to determine impact of plant". As discussed above, OTO-ZZ-00002 is not the correct procedure to implement.

B. Incorrect - OTO-ZZ-00002, Control Room Operations with Fire, Attachment B, Pre-Evacuation Contingency Actions is plausible as it is step #8 of OTO-ZZ-00002 but incorrect as this is not correct procedure to implement. Step #8 states "PERFORM Contingency Actions In Attachment B, Pre-Evacuation Contingency Actions, As Time And Personnel Permit" C. Correct - At Step #8 of OTO-ZZ-00001, the CRS is to "DIRECT Each Watchstander To Perform The Assigned Attachment" and the 2nd bullet states " Balance of Plant Operator (BOP)

Attachment B, Balance Of Plant (BOP) Operator Actions With Fire" This is correct as the BOP is one of the control room operators.

D. Incorrect - At Step #8 of OTO-ZZ-00001, the CRS is to "DIRECT Each Watchstander To Perform The Assigned Attachment" and the 3rd bullet states "Reactor Operator (RO)

Attachment C, Reactor Operator (RO) Actions With Fire". When the RO performs Attachment C.

and reaches Step #C10 "VERIFY DG NE02 - RUNNING" and if it isnt the RNO applies which is "

Go To Attachment F, Manual Start Of DG NE02 And Energizing NB02". This makes this choice plausible as it is included in the procedure for a control room operator actions but incorrect as it is a contingency procedure that is not initially directed by the CRS, it will be up to the discretion of the RO based on plant status. Additionally, there is no stem information that indicated the B EDG has or will fail therefore this contingency is not required but still remains plausible as it may be believed that the B EDG is started regardless during control room inaccessibility as it is the safe shutdown train.

Technical Reference(s):

1. OTO-ZZ-00001, Control Room Inaccessibility, Rev 048

2. OTO-KC-00001, Addendum C-27, Control Building 2047' Main Control Room, Rev 0

3. OTO-KC-00001, Addendum A-33, Aux Building 2026'-Aux Shutdown Panel Room 'B', Rev 0

4. OTO-ZZ-00002, Control Room Operations with Fire, Rev 9 References to be provided to applicants during examination: None Learning Objective: T61.0003B, LP 31, Objective E, Given a set of plant conditions or parameters 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 _____N/A_______

NRC Written Examination Callaway Plant Senior Reactor Operator Question Cognitive Level:

Memory or Fundamental Knowledge ____

Comprehension or Analysis __X___

LOD _____3_____

10 CFR Part 55 Content:

Comments:

k/a match as the choice (determine and interpret the need) for an emergency plant shutdown is the driving factor for the correct procedure attachment selection. Stated differently, the applicant must determine the need first and then select the appropriate attachment and by examining which attachment to implement it 1) meets the K/A and 2) makes it SRO ONLY per 10 CFR 55.43(b)(5).

SRO ONLY due to ES401 Figure 2 of NUREG 1021 as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely by knowing immediate operator actions? NO Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? NO Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 1 W/E08 RCS Overcooling - PTS Group # 2 K/A # W/E08 EA2.2 Importance Rating 4.1 Ability to determine and interpret the following as they apply to the (Pressurized Thermal Shock):

Adherence to appropriate procedures and operation within the limitations in the facilitys license and amendments.

REFERENCE PROVIDED Question #85 The crew has just entered to E-1, Loss of Reactor or Secondary Coolant, from E-0 and the following plant conditions exist:

All RCS Cold Leg temperatures are 170°F and lowering slowly

RCS pressure is 50 psig and stable

RHR flow indicates 6000 GPM

PZR level is off scale high (1) What is the status of the RCS INTEGRITY Critical Safety Function Status Tree?

And (2) Based on these conditions and after the crew has entered FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, what is the NEXT procedure the CRS should direct?

A. (1) An ORANGE Path exists (2) E-1, Loss of Reactor or Secondary Coolant B. (1) An ORANGE Path exists (2) EOP Addendum 5, Establishing Excess Letdown C. (1) A RED Path exists (2) E-1, Loss of Reactor or Secondary Coolant D. (1) A RED Path exists (2) EOP Addendum 5, Establishing Excess Letdown Answer: C

NRC Written Examination Callaway Plant Senior Reactor Operator Explanation: When RCS Pressure and Temperature are plotted on Figure 4a of CSF-1 (page 7 of 11), the values are to the Left of the Limit A Curve When checking the severity of the Integrity CSF (Figure 4 of CSF-1, page 6 of 11) Answer NO to the question Temperature Reduction in ALL RCS Cold Leg Less than 100F in 60 minutes and Answer NO to question ALL RCS Pressure VS Cold Leg Temperature Points to the Right of Limit A. This results in a RED Path.

For step #1 of FR-P.1, Check RCS Pressure - Greater Than 325 psig, the answer is NO. The RNO states "IF either RHR pump flow is greater than 850 GPM, THEN Return To procedure and step in effect". Therefore RHR flow is left as is and a Transition back to E-1 is performed.

EOP Addendum 5 is plausible as it is an option to address the high PZR level and directed by FR-P.1 Step 21 which states "CHECK PZR Level - LESS THAN 74% [64%]" and the RNO states

" CONTROL charging and letdown as necessary. IF PZR level can NOT be lowered with charging and letdown, THEN also ESTABLISH excess letdown using EOP Addendum 5, Establishing Excess Letdown."

A. Incorrect - a RED path exists B. Incorrect - both are wrong as a RED path exists and the procedure selection is not the next procedure to implement. Plausible if it is believed the crew should stay in FR-P.1 and at procedure step #21 establish excess letdown due to the high PZR level.

C. Correct - see above D. Incorrect - wrong procedure choice as explained above Technical Reference(s):

1. CSF-1, Critical Safety Function Status Trees (CSFST), Rev 13

2. FR-P.1, Response to Imminent Pressurized Thermal Shock Condition, Rev 14 References to be provided to applicants during examination:

1. CSF-1, Figure 4a, Limit A curve Learning Objective: T61.003D, Emergency Operations, LP #28 Objective G STATE and EXPLAIN the parameters which are evaluated, including their Criteria and Basis, to transition from the following procedures to another procedure.

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

Question Source: Bank # _ _____

Modified Bank # ____X__L16706______

New _______

Question History: Last NRC Exam ______

NRC Written Examination Callaway Plant Senior Reactor Operator Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

LOD _____3_____

10 CFR Part 55 Content:

Comments:

While the RED / ORANGE path entry requirements are RO knowledge, the subsequent procedure transition and associated administrative requirements for this transition in the Functional Restoration Procedures (FRP) are SRO knowledge. Entering E-1 (for a second time0 from a FRP is not considered the first and major EOP entry and hence an SRO Level Question SRO ONLY due to ES401 Figure 2 of NUREG 1021 as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely by knowing immediate operator actions? NO Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? NO Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures-

Rev. 013 CSF-1 CRITICAL SAFETY FUNCTION STATUS TREES CONTINUOUS USE (CSFST) Page 7 of 11 Figure 4a Limit A Curve

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 Engineered Safety Features Actuation Group # 1 System K/A # 013 A2.04 Importance Rating 4.2 Ability to (a) predict the impacts of the following malfunctions or operations on the ESFAS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Loss of instrument bus.

Question #86 Reactor Power is 100%.

A ESFAS status panel audible alarm is received and the RO immediately recognizes that BG-LI-185 failed low. NCP suction has shifted from the VCT to the RWST

One SG NR level indication fails low on each SG (1) Based on these indications the MDAFW pumps _____(1)_____ receive a start signal from the ESFAS system.

And (2) Based on the above conditions, what is the FIRST procedure the CRS should direct?

A. (1) will (2) OTO-NN-00001, Loss of Safety Related Instrument Power, Attachment P, Loss of NN04 B. (1) will (2) OTO-BG-00004, VCT Level Channel Failures, Attachment C, BG-LT-185 Failure C. (1) will NOT (2) OTO-NN-00001, Loss of Safety Related Instrument Power, Attachment P, Loss of NN04 D. (1) will NOT (2) OTO-BG-00004, VCT Level Channel Failures, Attachment C, BG-LT-185 Failure Answer: C

NRC Written Examination Callaway Plant Senior Reactor Operator Explanation:

The SG level indications in the stem indicate that it could be a loss of any NN bus. The VCT swapover to the RWST indicates that it is either NN01 or NN04 due to the power supplies of the VCT level indicators (BG-LI-112 or 185).

Per sheet 7250D64 Sheet 15, AFW Pump startup, ESFAS logic requires 2 of 4 channels on 1 of 4 SG to start the MDAFW pump. The indication only show 1 channel failing on each SG, therefore the MDAFP will NOT receive an automatic start signal from ESFAS.

While both procedure attachments have actions to mitigate the boration event (swapover to the RWST), the correct procedure to prioritize is OTO-NN-00001 Attachment P as its action will stabilize the entire plant (including the BB LI 185 failure) vice the specific issue of the instrument.

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

1. OTO-NN-00001, Loss of Safety Related Instrument Power, Rev 38

2. 7250D64 Sheet 15, AFW Pump Startup, Rev 7

3. OTO-BG-00004, VCT Level Channel Failures, Rev 20 References to be provided to applicants during examination: None Learning Objective: T61.003B. Off Normal Operations, LP#-27, Objective C: Given a set of plant conditions or parameters indicating a Loss of Safety Related Instrument Power, ANALYZE the correct procedure(s) to be utilized and the required actions to stabilize the plant.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam _____N/A______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____2____

10 CFR Part 55 Content:

Comments:

NRC Written Examination Callaway Plant Senior Reactor Operator SRO Only as the question requires the applicant to prioritize procedure attachments and understand the Off normal procedure network and its content and overall intent and function of separate off normal procedures.

SRO ONLY due to ES401 Figure 2 of NUREG 1021 as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely by knowing immediate operator actions? NO Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? NO Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps? NO

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures? NO

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures? NO

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 Containment Cooling Group # 1 K/A # 022 G2.4.21 Importance Rating 4.6 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc.

Question #87 Following a large break LOCA, current plants conditions are:

NB01 is locked out

Containment pressure is 30 psig

Containment radiation is 2.8 R/hr (1) What YELLOW path should the CRS enter?

And (2) Per ODP-ZZ-00025, EOP/OTO Users Guide, and given these plant conditions, how often should CSFSTs be monitored?

A. 1) FR-Z.1, Response to High Containment Pressure

2) Continuously B. 1) FR-Z.1, Response to High Containment Pressure

2) 15 minutes C. 1) FR-Z.3, Response to High Containment Radiation Level

2) Continuously D. 1) FR-Z.3, Response to High Containment Radiation Level

2) 15 minutes Answer: B Explanation: Per CSF-1 figure 5 with the given conditions entry into FR-Z.1 would be warranted, entry into FR-Z.3 would not be met until containment rad is 3 rad/hr. Per ODP-ZZ-00025, step 4.24.10, Red and Orange paths should be monitored continuously and Yellow paths should be monitored every 10 to 20 minutes A. Incorrect - Yellow paths should be monitored every 10 to 20 minutes B. Correct

NRC Written Examination Callaway Plant Senior Reactor Operator C. Incorrect - Containment rad not high enough to enter FR-Z.3 and Red and Orange paths are monitored continuously D. Incorrect - Containment rad not high enough to enter FR-Z.3 Technical Reference(s):

1. CSF-1, Critical Safety Function status Trees, Rev 013

2. ODP-ZZ-00025, ROP/OTO User's Guide, Rev 038 References to be provided to applicants during examination: None Learning Objective: T61.003D, Emergency Operations, LP# CSF and Accident Demo, Enabling Objectives

Discuss use of CSF-1 as detailed in ODP-ZZ-00025, EOP/OTO Users Guide.

Analyze the methods to implement CSF-1.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____NA______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

LOD ____4______

10 CFR Part 55 Content:

Comments:

SRO ONLY due to ES401 Figure 2 of NUREG 1021 as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely by knowing immediate operator actions? NO Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? NO Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

NRC Written Examination Callaway Plant Senior Reactor Operator

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures-

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 AC Electrical Distribution Group # 1 K/A # 062 G2.2.37 Importance Rating 4.6 Ability to determine operability and/or availability of safety related equipment.

Question #88 Reactor Power is 100%.

Annunciator 28B, NN14 INV TRBL/XFR, has just alarmed

Local indication shows that the Bypass Source Supplying Load, P202, light is LIT

1) What is the status of NN14, 120V AC Vital Inverter?

And 2.) What Technical Specification action(s) will to be taken to mitigate the situation?

____(1)____ ____(2)____

A. INOPERABLE 3.8.7 Inverters Operating ONLY B. OPERABLE 3.8.7 Inverters Operating ONLY C. INOPERABLE 3.8.7 Inverters Operating AND 3.8.9 Distribution Systems Operating D. OPERABLE 3.8.7 Inverters Operating AND 3.8.9 Distribution Systems Operating Answer: A Explanation: Per T/S 3.8.7 bases, in order to be considered operable the inverters be powered from a 125 VDC station battery, and supply power to the associated Vital bus.

NRC Written Examination Callaway Plant Senior Reactor Operator A. Correct -NN14 is INOPERABLE until it is supplying the 120V AC Vital Bus NN04 from the battery. NN04 will remain OPERABLE if it is energized from the bypass source per T/S 3.8.9.

T/S 3.8.7 for the INOPERABLE inverter is the only T/S entered.

B. Incorrect - The inverter NN14 is INOPERABLE, however NN04 will remain OPERABLE if it is energized from the bypass source. T/S 3.8.7 for the INOPERABLE inverter is the only T/S entered.

C. Incorrect - NN14 is INOPERABLE until it is supplying the 120V AC Vital Bus NN04 from the battery. NN04 will remain OPERABLE if it is energized from the bypass source per T/S 3.8.9.

T/S 3.8.7 for the INOPERABLE inverter is the only T/S entered. T/S 3.8.9 is referenced from T/S 3.8.7 to be entered ONLY if NN04 is deenergized. In this case NN04 is energized from the bypass source.

D. Incorrect - NN14 is INOPERABLE until it is supplying the 120V AC Vital Bus NN04. NN04 will remain OPERABLE if it is energized from the bypass source. T/S 3.8.7 for the INOPERABLE inverter is the only T/S entered. T/S 3.8.9 is referenced from T/S 3.8.7 to be entered if NN04 is deenergized. In this case NN04 is energized from the bypass source.

Technical Reference(s):

1. Technical Specification 3.8.7 Inverters - Operating References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#6, Objective G: EXPLAIN the Technical Specifications and bases for the Safeguards Power System.

Question Source: Bank # _No ID___

Modified Bank # ______

New _______

Question History: Last NRC Exam ___2014_____

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis _X___

LOD ____2______

10 CFR Part 55 Content:

Comments:

SRO Only due to 43(b) #2 - Facility operating limitations in the technical specifications and their bases as the information required to determine its operability is located in the TS Bases.

NRC Written Examination Callaway Plant Senior Reactor Operator Additionally per Figure 2-1 Attachment 2 of ES-401,

Can question be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS/TRM Action? Is No

Can question be answered solely by knowing the LCO/TRM information listed above-the-line? is No

Can question be answered solely by knowing the TS Safety Limits? Is No

Does the question involve one or more of the following for TS, TRM, or ODCM?

Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1) knowledge of TS bases that is required to analyze TS-required actions and terminology is Yes which means this is an SRO Only question

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 Emergency Diesel Generator 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 The plant is in Mode 4.

A Emergency Diesel Generator (EDG), NE01, is in a post maintenance run

It has run for 26 hours3.009259e-4 days 0.00722 hours 4.298942e-5 weeks 9.893e-6 months at 1.15 MW

Maintenance has requested that the Control Room secure the D/G (1) What describes the impact on NE01?

And (2) What action(s) should the CRS direct NEXT?

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, Section 5.6, Diesel Generator A Shutdown B. 1) Lube oil temperature will be low

2) Lower load on NE01 to 0.2 MW and secure the EDG IAW OTN-NE-0001A, Section 5.6, Diesel Generator A Shutdown C. 1) Oil will build up in the exhaust

2) Raise load to > 50% for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months IAW OTN-NE-0001A, Section 5.3, Paralleling Diesel Generator A to XNB01 D. 1) Lube oil temperature will be low

2) Raise load to > 50% for 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months IAW OTN-NE-0001A, Section 5.3, Paralleling Diesel Generator A to XNB01 Answer: C Explanation:

NRC Written Examination Callaway Plant Senior Reactor Operator OTN-NE-0001A 3.1.5 states:

If necessary to operate the engine with less than 20 % load (1.24 MWe) for greater than 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months , the engine should be run at greater than 50 % load (3.1 MWe) for at least one (1) hour in each 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months period, to reduce accumulation of combustion and lubrication products in the exhaust system.

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):

1. OTN-NE-0001A, Standby Diesel Generation System - Train A, Rev 052 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#3, Objective 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___L17687_

Modified Bank # ______

New _______

Question History: Last NRC Exam ____2013_____

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

LOD _____2_____

10 CFR Part 55 Content:

Comments:

SRO ONLY due to ES401 Figure 2 of NUREG 1021 as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely by knowing immediate operator actions? NO

NRC Written Examination Callaway Plant Senior Reactor Operator Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? NO Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures-

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 Process Radiation Monitoring Group # 1 K/A # 073 A2.01 Importance Rating 2.9 Ability to (a) predict the impacts of the following malfunctions or operations on the PRM system; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Erratic or failed power supply Question #90 Reactor Power is 100% with SFP fuel movement in progress.

A power supply issue on Fuel Building Radiation, GG RE-27 gas channel causes a HiHi alarm.

What procedure(s) should the CRS direct?

A. OTO-KE-00001, Attachment A, SA066X CRVIS Train A Verification ONLY B. OTO-SA-00001, Attachment AB, CRVIS Train A Verification ONLY C. OTO-KE-00001, Attachment A, SA066X CRVIS Train A Verification AND Attachment B, SA066Y CRVIS Train B Verification D. OTO-SA-00001, Attachment AB, CRVIS Train A Verification, AND Attachment AC, CRVIS Train B Verification Answer: D Explanation:

A HiHi reading on GGRE27 will cause both trains of FBIS and both train of CRVIS to actuate.

OTO-KE-00001 is plausible as a distractor as GG RE 27 abnormal reading is an entry condition to the procedure. Attachment A and B could be implement from a RNO procedure step if not all of the CRVIS components repositioned as designed (no indication of that in the stem).

All Choices were limited to CRVIS (i.e no FBIS) to prevent non plausible distractors.

A. Incorrect -See above explanations. If it is believed that this A Train detector will only cause an A Train CRIVS and FBIS, then attachment A is a plausible choice.

B. Incorrect -See above explanations. If it is believed that this A Train detector will only cause an A Train CRIVS and FBIS, then attachment A is a plausible choice.

C. Incorrect -See above explanations D. Correct - This is the correct procedure and applicable Attachments for both trains CRVIS actuations in this plant condition.

NRC Written Examination Callaway Plant Senior Reactor Operator Technical Reference(s):

1. OTO-SA-00001, ESFAS Verification And Restoration, Rev 042

2. OTO-KE-00001, Fuel Handling Accident, Rev 16 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#36, Objective C: IDENTIFY the Process and Area Radiation Monitoring Control Room controls, alarms, and indications and DESCRIBE how each is used to predict, monitor and control the Process and Area Radiation Monitoring System.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X__

LOD _____3_____

10 CFR Part 55 Content:

Comments:

The question is written the b part of the K/A which is allowed per NUREG 1021 as knowledge of the K/A part A is required to correct answer part B.

SRO ONLY due to ES401 Figure 2 of NUREG 1021 as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely by knowing immediate operator actions? NO Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? NO Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

NRC Written Examination Callaway Plant Senior Reactor Operator

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

Knowledge of when to implement attachments and appendices, including how to coordinate these items with procedure steps

Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures

Knowledge of administrative procedures that specify hierarchy, implementation, and/or coordination of plant normal, abnormal, and emergency procedures-

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 Fuel Handling equipment Group # 2 K/A # 034 A2.03 Importance Rating 4.0 Ability to (a) predict the impacts of the following malfunctions or operations on the Fuel Handling System ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Mispositioned fuel element Question #91 The plant is in Mode 6.

During core mapping post core load, Reactor Engineering reports that two fuel bundles are in the wrong location

RWST level is 16%

Audible and Source Range counts began to rise on both SR NIs

Annunciator 65A, SR HIGH FLUX AT SHUTDOWN, alarms

Reactor Makeup Control System (RMCS) is NOT available (1) What action should the CRS direct NEXT?

And (2) If SRNIs counts continue to rise and criticality occurs, what is the MAXIMUM time allowed to notify the NRC?

A. 1) Emergency borate

2) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months B. 1) Emergency borate

2) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C. 1) Actuate Containment Purge Isolation

2) 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months D. 1) Actuate Containment Purge Isolation

2) 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Answer: A Explanation: OTA-RK-00020 Addendum 65A has three potential procedures the CRS can use:

OTO-KE-00001, OTO-ZZ-00003, and OTO-SE-00001. Based on the given conditions the correct procedure is OTO-ZZ-00003. The NEXT action is to Emergency borate per step 7 RNO (the RMCS is unavailable). The one hour notification is required for Accidental Criticality per APA-ZZ-

NRC Written Examination Callaway Plant Senior Reactor Operator 000520 Attachment 2.

Actuate Containment Purge Isolation is step #13 of OTO-KE-00001, Fuel Handling accident and is plausible for 2 reasons - 1) if they believe the fuel handling accident procedure applies for this instance of misloaded fuel assemblies and 2) as the plant is in mode 6 (head removed) it is plausible to think that the action to isolate containment should be the NEXT action.

A. Correct - see above explanation B. Incorrect - see above explanation - this is a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months notification C. Incorrect - see above explanation D. Incorrect - see above explanation Technical Reference(s):

1. OTA-RK-00020 Addendum 65A, Rev 000

2. OTO-ZZ-00003, Loss of Shutdown Margin / Dilution Event, Rev 018

3. APA-ZZ-000520, Reporting Requirements and Responsibilities, Rev 054

4. OTO-KE-00001, Fuel Handling Accident, Rev 16 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP#33, Objective E. Given a set of plant conditions or parameters indicating a Loss of Shutdown Margin, IDENTIFY the correct procedure(s) to be utilized and OUTLINE the high level actions to stabilize the plant.

T61.003A, Normal Operations, LP#31 Event Review and Reportability, Objective B: PERFORM the following as they pertain to APA-ZZ-00520, Reporting Requirements and Responsibilities

DISCUSS the incidents reportable in the following time frames:

o 1 Hour Question Source: Bank # _X_L23233__

Modified Bank # ______

New _______

Question History: Last NRC Exam ___NA_______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____4_____

10 CFR Part 55 Content:

Comments:

SRO per criteria 1 due to the reporting requirements associated with the facility license.

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 Area Radiation monitoring Group # 2 K/A # 072 A2.02 Importance Rating 2.9 Ability to (a) predict the impacts of the following malfunctions or operations on the ARM system-and (b) based on those predictions, use procedures to correct, control, or mitigate the onsequences of those malfunctions or operations: Detector failure Question #92 New fuel movement is in progress in the new fuel storage area when:

Annunciator 62C, Area Radiation Monitor Failure, alarms

RO reports SDRE0036, NEW FUEL STOR AREA RAD, reads LOW (1) Per the FSAR, with SDRE0036 failed LOW, can new fuel movement continue?

And (2) If SDRE0035, NEW FUEL STOR AREA RAD, also fails LOW, what FSAR action would be required to support new fuel movement?

A. 1) No

2) Restore BOTH channels to functional status within 30 days B. 1) Yes

2) Restore BOTH channels to functional status within 30 days C. 1) No

2) Install a portable continuous monitor with the same alarm setpoint to allow for up to 30 days of fuel movement D. 1) Yes

2) Install a portable continuous monitor with the same alarm setpoint to allow for up to 30 days of fuel movement Answer: D Explanation: FSAR 16.3.3.7 With no channel FUNCTIONAL, operation may continue for up to 30 days provided an appropriate portable continuous monitor with the same alarm setpoint is provided in the new fuel storage area. Restore one channel to FUNCTIONAL status within 30 days or suspend all operations involving fuel movement in the fuel building.

NRC Written Examination Callaway Plant Senior Reactor Operator A. Incorrect - both parts are incorrect - new fuel is allowed with one channel functional and ONLY restore one channel to FUNCTIONAL status within 30 days or suspend fuel movement in the fuel building B. Incorrect - part 2 is incorrect - Need to restore ONLY one channel to FUNCTIONAL status within 30 days or suspend fuel movement in the fuel building.

C. Incorrect - part 1 is incorrect - new fuel is allowed with ONLY one channel functional D. Correct see above explanation Technical Reference(s):

1. OTA-RK-00020, Addendum 62C, Area Radiation Monitor Failure, Rev 001

2. FSAR 16.3.3.7 New Fuel Storage Area Criticality Monitor Limiting Condition for Operation References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP-36, Objective G: LIST the Process and Area Radiation Monitors required by Technical Specifications and the Final Safety Analysis Report (FSAR) Chapter 16.

Question Source: Bank # ______

Modified Bank # ______

New ____X__

Question History: Last NRC Exam _____NA_____

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD _____3_____

10 CFR Part 55 Content:

Comments:

SRO Only due to 43(b) #2 - Facility operating limitations in the technical specifications and their bases.

Additionally per Figure 2-1 Attachment 2 of ES-401,

Can question be answered solely by knowing 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months TS/TRM Action? Is No

NRC Written Examination Callaway Plant Senior Reactor Operator

Can question be answered solely by knowing the LCO/TRM information listed above-the-line? is No

Can question be answered solely by knowing the TS Safety Limits? Is No

Does the question involve one or more of the following for TS, TRM, or ODCM?

Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1) knowledge of TS bases that is required to analyze TS-required actions and terminology is Yes which means this is an SRO Only question

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 2 Fire Protection Group # 2 K/A # 086 G2.4.41 Importance Rating 4.6 Knowledge of the emergency action level thresholds and classifications.

REFERENCE PROVIDED Question #93

@0800 Reactor power is 100% when the control room received multiple reports (Control Room KC008 panel and field) of a fire in NB01.

@0815 NB01 is locked out. Fire fighting is in progress.

@0820 Offsite fire fighting support has been requested by the control room.

@0830 Offsite support arrived and began assisting Callaway personnel.

@0845 The fire spreads and damages NE02, 'B' Train EDG. NB0211, NB02 Emergency Supply Breaker, lost all control room indication.

@0900 Field reports state that the fire is out.

(1) Based on the above events, what is the LATEST time the FIRST notification to the NRC Operations Center should be made?

And (2) What is the HIGHEST Emergency Action Level (EAL) associated with these events?

A. (1) 0915 (2) Unusual Event B. (1) 1000 (2) Unusual Event C. (1) 0915 (2) Alert D. (1) 1000 (2) Alert

NRC Written Examination Callaway Plant Senior Reactor Operator Answer: C Explanation:

Notification of the state and county is 15 minutes and notification to the WHO i.e. the NRC operation center is 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months (per step 5.3.4 of EIP-ZZ-00102) after the EAL declaration.

There are 2 EALs associated with the Fire as a hazardous event. HU4.1 and SA9.1.

HU4.1 A FIRE is not extinguished within 15 min. of any of the following FIRE detection indications: (Note 1)

Report from the field (i.e., visual observation).

Receipt of multiple (more than 1) fire alarms or indications.

Field verification of a single fire alarm.

AND The FIRE is located within any Table H-1 area.

SA9.1 The occurrence of any Table S-5 hazardous event.

AND Event damage has caused indication of degraded performance on one train of a SAFETY SYSTEM needed for the current operating MODE AND EITHER:

Event damage has caused indications of degraded performance in a second train of a SAFETY SYSTEM needed for the current operating MODE.

Event damage has resulted in VISIBLE DAMAGE to a second train of a SAFETY SYSTEM needed for the current operating MODE.

Note: SU1.1 does not apply as offsite power was never lost to NB02 during these events.

@0800 conditions exists for HU4.1 and this is the first EAL reached during the series of events but not the highest. The EAL must be declared by 0815 and therefore the NRC operations Center must be notified by 0915 which is the LATEST time for the FIRST notification.

@0845 conditions exist for SA9.1 (loss of NB01 and Event damage has caused indications of degraded performance in the second train specifically NE02 and associated output breaker) and is the Highest EAL reached during the event. This EAL must be declared by 0900 and therefore the NRC operations center must be notified by 1000. However, this is not the first required notification time and therefore incorrect. 1000 is also plausible is it may be believed that a notification is required 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the fire is out (@0900).

If the damage to B Train Safety Systems is not accounted for, the highest EAL would be HU4.1, an unusual event, making it a plausible distractor.

A. Incorrect - See above explanation - the EAL level is not the highest reached during the series of events B. Incorrect - See above explanation - both are wrong.

C. Correct D. Incorrect - See above explanation - wrong notification time as discussed above Technical Reference(s):

1. EIP-ZZ-00101 Addendum 1 EAL classification matrix, Rev 10

NRC Written Examination Callaway Plant Senior Reactor Operator

2. EIP-ZZ-00102, Emergency Implementing Actions, Rev 65 References to be provided to applicants during examination:

1. EIP-ZZ-00101 Addendum 1 Wallchart EAL classification matrix, Rev 10 Learning Objective: T68.1020 RERP, LP#1, Objective B: Determine the emergency classification for given indications and/or symptoms, per EIP-ZZ-00101.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ______N/A______

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

LOD ____3____

10 CFR Part 55 Content:

Comments:

SRO only due to (5) - Assessment of facility conditions and selection of appropriate procedures as follows:

Can the question be answered solely by knowing systems knowledge, i.e., how the system works, flowpath, logic, component location? NO Can the question be answered solely by knowing immediate operator actions? NO Can the question be answered solely by knowing entry conditions for AOPs or plant parameters that require direct entry to major EOPs? NO Can the question be answered solely by knowing the purpose, overall sequence of events, or overall mitigative strategy of a procedure? NO Does the question require one or more of the following?

Assessing plant conditions (normal, abnormal, or emergency) and then selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed YES

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Conduct of Operations Group # N/A K/A # G2.1.34 Importance Rating 3.5 Knowledge of primary and secondary plant chemistry limits.

REFERENCE PROVIDED Question #94 Reactor Power is 100%.

The following chemistry results have been reported to the control room:

RCS Chloride = 65 ppb

RCS Sulfate = 80 ppb

RCS Fluoride = 20 ppb

RCS Dissolved Oxygen = 10 ppb

SG Feedwater Oxygen = 15 ppb

Condensate Oxygen = 80 ppb

'A' and 'C' SG Chloride = 40 ppb

'B' SG Chloride = 35 ppb

'D' SG Chloride = 45 ppb

All SG Sodium = 30 ppb (1) What is the HIGHEST plant (Primary and/or Secondary) chemistry action level exceeded?

And (2) Who may approve plant operation outside of the guidelines for the action level?

A. (1) Action Level 1 (2) Senior Director, Nuclear Operations B. (1) Action Level 2 (2) Senior Director, Nuclear Operations C. (1) Action Level 1 (2) Shift Manager D. (1) Action Level 2

NRC Written Examination Callaway Plant Senior Reactor Operator (2) Shift Manager Answer: B Explanation:

Per APA-ZZ-01020 and APA-ZZ-01021section 3.0, the senior director, nuclear operations, duty manager, and chemistry manager all "Approves operation of the plant outside guidelines of this procedure for all Action Levels." Per Step 3.3 the shift manager "With Chemistry Management, interprets critical plant data and designates specific plant maneuvers to prevent exceeding Action Levels."

Per Attachment 1 or APA-ZZ-01020 and APA-ZZ-01021 ADD 1, all chemistry parameters are in action level 1 status except for SG Feedwater Oxygen Chloride level which is action level 2.

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

1. APA-ZZ-01020, Primary Chemistry Program, Rev 32

2. APA-ZZ-01021, Secondary Chemistry Program, Rev 35

3. APA-ZZ-01021, ADD 1, Action Level Limits Part I and II, Rev 7 References to be provided to applicants during examination:

1. APA-ZZ-01020, Primary Chemistry Program, Rev 32, Attachment 1 (edited)

2. APA-ZZ-01021, ADD 1, Action Level Limits Part I and II, Rev 7 (edited)

Learning Objective: T61.003A, LP #22, Objective B and C:

B. DISCUSS the following as they pertain to APA-ZZ-01020, Primary Chemistry Program:

1. Purpose and Scope

2. Responsibilities of:

a. Plant Director

b. EDO

c. Shift Manager

3. Required Actions for:

a. Action Levels Exceeded

b. Action Level 1 Exceeded

c. Action Level 2 Exceeded

d. Action Level 3 Exceeded

4. Definitions of:

a. Action Levels

b. Action Level 1

c. Action Level 2

d. Action Level 3

e. Status Modes C. DISCUSS the following as they pertain to APA-ZZ-01021, Secondary Chemistry

NRC Written Examination Callaway Plant Senior Reactor Operator Program:

1. Purpose and Scope

2. Responsibilities of:

a. Senior Director, Nuclear Operations

b. EDO

c. Shift Manager

3. Required Actions for:

a. Actions Common to All Action Levels

b. Action Level 1 Exceeded

c. Action Level 2 Exceeded

d. Action Level 3 Exceeded

4. Requirements for MODE changes

5. Definitions of:

a. Status Modes

b. Action Levels Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis ___X__

LOD ____3____

10 CFR Part 55 Content:

Comments:

k/a match as knowledge of the limits and responsibilities when a limit are exceeded met the intent of the k/a of "knowledge of the primary and secondary chemistry limit" SRO only per 10 CFR55.43b(1) - conditions and limitations of the facility License. Maintaining plant chemistry within standards and the actions required when outside the limit is apart of the conditions and limitations of the license.

APA-ZZ-01020 Rev. 032 Attachment 1 Sheet 1 of 2 ACTION LEVELS Table 1: Reactor Coolant System Power Operation (MODES 1 and 2 (Reactor Critical))

Action Level Parameter 1 2 3 Chloride, ppb

>50 >150 >1500 Fluoride, ppb

>50 >150 >1500 Sulfate, ppb

>50 >150 >1500 Dissolved Oxygen, ppb(e)

>5 >100 >1000

<25 Hydrogen, cc/kg

>50 <15 <5 Lithium, ppm NA NA Page 16 of 59 INFORMATION USE

APA-ZZ-01021, ADDENDUM 1 Rev. 007 Part I:

Action Levels at >50% Reactor Power Steam Generator Action Level (AL)

Parameter 1 2 3 Cation Conductivity, µS/cm N/A Sodium, ppb >5 >50 >250 Chloride, ppb >10 >50 >250 Sulfate, ppb >10 >50 >250 Condensate Action Level (AL)

Parameter 1 2 3 Oxygen , ppb 100 N/A N/A Page 2 of 7 INFORMATION USE

APA-ZZ-01021, ADDENDUM 1 Rev. 007 Part I: Action Levels at >50% Reactor Power (continued)

SG Feedwater Action Level (AL)

Parameter 1 2 3 ETA, ppm <1.0 N/A N/A MPA, ppm <3.0 N/A N/A Ammonia, ppm <3.0 N/A N/A Oxygen, ppb >5 >10 N/A Total Iron, ppb >5 N/A N/A Total Copper, ppb >1 N/A N/A Hydrazine, ppb < 8 X CPD DO2 OR < 20 ppb (f) N/A Hydrazine to oxygen ratio <2 for >8 hours (b)

Page 3 of 7 INFORMATION USE

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Conduct of Operations Group # N/A K/A # G2.1.41 Importance Rating 3.7 Knowledge of the refueling process.

Question #95 You have been directed to perform an AUTO TRANSFER of a fuel assembly from the Fuel Building to Containment.

What conditions are required to initiate an auto transfer sequence?

A. Both upender frames must be full DOWN and the refuel machine is clear of the upender in Containment.

B. Both upender frames must be full UP and the SFP Bridge Crane is clear of the upender in the Fuel Building.

C. Rx Building upender frame must be full UP and Fuel Building upender frame must be full DOWN. Refuel machine and SFP Bridge Crane must be clear of the upenders.

D. Fuel Building upender frame must be full UP and Rx Building upender frame must be full DOWN. Refuel machine and SFP Bridge Crane must be clear of the upenders.

Answer: D Explanation:

OTS-KE-00015 Section 5.7 explains the process for Auto Fuel transfer (from the FB to CTMT) and is summarized as follows:

5.7.3. ENSURE the FB side upender is up, 5.7.4. CHECK the SFP Bridge Crane is clear of the upender 5.7.6. CHECK the RX SIDE FRAME DOWN 5.7.8. ENSURE that an upender clear condition exists at the RB side.

A. Incorrect as the FB upender is required to be up not down as the distractor indicates.

Plausible is the entire transfer process with associated interlocks are not known or understood.

B. Incorrect as the RB upender is required to be down not up as the distractor indicates.

Plausible is the entire transfer process with associated interlocks are not known or understood.

C. Incorrect as the RB and FB upender frames in this distractor are opposite as required (i.e the frames are in position to transfer from the CTMT to FB but this is not the direct the stem. The machine and crane interlocks are in the correct positions for this transfer. Plausible is the entire

NRC Written Examination Callaway Plant Senior Reactor Operator transfer process with associated interlocks are not known or understood or the direct of fuel movement and associated positions are confused.

D. Correct - See explanation above Technical Reference(s):

1. OTS-KE-00015, Fuel Transfer System, Rev 29 References to be provided to applicants during examination: None Learning Objective: T61.0110, Systems, LP#79, Objective H: Describe the interlocks and protective features of the following:

1. New fuel elevator

2. Spent fuel bridge crane

3. Transfer system

4. Refueling machine gripper Question Source: Bank # __X R12315____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____4____

10 CFR Part 55 Content:

10 CFR 55.43(b)(7)

Comments:

SRO Only due to 43(b)(7) - Fuel handling facilities and procedures

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Equipment Control Group # N/A K/A # G2.2.5 Importance Rating 3.2 Knowledge of the process for making design or operating changes to the facility.

Question #96 Reactor Power is 100%.

A safety related component is experiencing RCS in-leakage and requires draining twice per shift

A work package has been developed to isolate the component by closing a normally open valve

The valve control circuitry will be modified to receive an open signal such that the safety related component could perform its design purpose

The safety related component and the normally open valve's control circuitry will be returned to original design configuration during the next refueling outage which is scheduled 79 days from now What describes this modification and the correct process to implement it?

A. It may affect the safety related function; process this work package through the 10CFR50.59 evaluation process B. It establishes plant conditions that are in compliance with Technical Specifications; only a 10CFR50.59 screening is required C. It meets the same functional requirements as the original design; the work package can be approved for implementation by performing only a 10CFR50.59 screening D. It will be in service for fewer than 90 days, process this work package through the temporary modification process without performing a 10CFR50.59 evaluation Answer: A Explanation:

This is a change to a safety related SSC and will follow the design change process and a 50.59 will be required.

Note: Temporary Modifications are temporary changes to design configuration of the plant, which would be considered as Engineering Changes if made on a permanent basis.

NRC Written Examination Callaway Plant Senior Reactor Operator A. Correct B. Incorrect - plausible if the candidate does not determine that this a change to the normal system lineup and system design.

C. Incorrect - plausible if the candidate recognizes that this is a change to the normal full power lineup (i.e. outlet valve closed but "meets the same functional requirements .") but believes that since it performs the same design function only a screening is required D. Incorrect - plausible if the candidate believes that this follows / fits into the category of temporary modifications as it will be removed next refuel outage and that this process does not require a 50.59 evaluation.

Technical Reference(s):

1. APA-ZZ-00600, Design Change Control, Rev 60

2. APA-ZZ-00605, Temporary System Modifications, Rev 37 References to be provided to applicants during examination: None Learning Objective: T61.003A, Normal Operations, LP #18, Objectives A and B:

A. PERFORM the following as it pertains to APA-ZZ-00600, Design Change Control:

1. EXPLAIN the purpose of the procedure

2. DESCRIBE the process for initiating a request for change

3. DISCUSS Concurrent Changes to include:

a. When they are used

b. Requirements for system to be out-of-service

c. Operations approval requirements B. PERFORM the following concerning APA-ZZ-00605, Temporary System Modifications:

1. EXPLAIN the purpose of the procedure

2. IDENTIFY the conditions that are excluded from this procedure

3. DISCUSS the difference in requirements for approval for non-safety and safety related equipment temporary modifications

4. EXPLAIN the Temporary Modification Audit process

5. DESCRIBE the following concerning Procedurally Controlled Temporary Modifications (PCTMs):

a. When they can be used

b. Limitations

c. Time limits for:

Question Source: Bank # __ X L16756____

Modified Bank # ______

New _______

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge _____

Comprehension or Analysis __X___

NRC Written Examination Callaway Plant Senior Reactor Operator LOD ____3____

10 CFR Part 55 Content:

10 CFR 55.43(b)(3)

Comments:

SRO ONLY due to 10 CFR 55.43(b)(3) - Facility licensee procedures required to obtain authority for design and operating changes in the facility.

Modified bank question to a generic non system nature but the met the definition of a modified question.

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Equipment Control Group # N/A K/A # G2.2.17 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 An unplanned equipment inoperability results in the entry into a Red Risk Condition.

(1) Per EDP-ZZ-01129, Callaway Energy Center Risk Assessment, operator actions to mitigate plant risk _____(1)_____ be delayed until a risk assessment is performed.

And (2) Per ODP-ZZ-00001, Addendum 13 Shift Manager Communications, the Shift Manager is required to notify the ____(2)______ of this Red Risk Condition.

A. (1) should (2) Senior Vice President-Nuclear B. (1) should (2) Director, Nuclear Operations C. (1) should NOT (2) Senior Vice President-Nuclear D. (1) should NOT (2) Director, Nuclear Operations Answer: D Explanation: Per ODP-ZZ-00001 ADD 13 Attachment 1, "Equipment inoperability that results in the unplanned entry of into a Red Condition per EDP-ZZ-01129" (page 6), requires the Shift Manager to notify Duty Manager, NRC RI, SDNO/DNO, SSS, and Duty Team. Senior Vice President Nuclear is plausible as he/she is listed on the matrix but not contacted by the shift manager for this situation. The Senior Vice President Nuclear would be notified when " High potential exists to have to shut down the unit due to Tech. Spec. action statement, equipment problem (non-Tech. Spec.), Chemistry limits, etc." for example.

Per EDP-ZZ-01129 step 4.3.6 states "Risk assessment for unplanned maintenance should be

NRC Written Examination Callaway Plant Senior Reactor Operator performed per this procedure, but should NOT delay operator and/or maintenance from taking timely actions to mitigate plant risk." Step 4.3.6.a states "Emergent work may require troubleshooting. Since these activities can result in equipment being removed from service or subjected to transient conditions, all troubleshooting on equipment within the 10 CFR 50.65 (a)(4) scope is performed using appropriate administrative controls and a risk assessment is performed prior to the troubleshooting being performed." Therefore, operator actions to mitigate the Red Risk conditions may be performed prior to the risk assessment being performed but troubleshooting activities will be delayed until the completion of the risk assessment making "Should" be delayed plausible but incorrect.

A. Incorrect - See explanation above - both are wrong B. Incorrect - See explanation above - the delay of action in part 1) is wrong C. Incorrect - See explanation above - notification of the Senior Vice President Nuclear is incorrect D. Correct Technical Reference(s):

1. ODP-ZZ-00002, Equipment Status Control, Rev 93

2. EDP-ZZ-01129, Callaway Energy Center Risk Assessment, Rev 50

3. ODP-ZZ-00001, ADD 13, Shift Manager Communications, Rev 24 References to be provided to applicants during examination: None Learning Objective: T61.003A, Normal Operations, LP #74, Objective B: DESCRIBE the following as it pertains to ODP ZZ 00002, 'Equipment Status Control":

1. The purpose and scope

2. The definition of:

a. Operable/Operability

b. Safety Related

3. Shift Manager responsibilities

4. The process for:

a. Initiating an EOSL entry

b. Clearing an EOSL entry

c. Maintaining the EOSL

5. The application of Operability Evaluations.

6. IMPLEMENTATION of Risk Management Actions required per ODP-ZZ-00002 Appendix.

Note: ODP-ZZ-00002 directs several actions IAW or per EDP-ZZ-01129 therefore the EDP procedure is included in the objectives scope as well as the responsibilities of the ODP-ZZ-00002 direct several notification responsibilities of the Shift Manger.

Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

NRC Written Examination Callaway Plant Senior Reactor Operator Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3____

10 CFR Part 55 Content:

10 CFR 55.43(b)(1) - Conditions and limitation in the facility license Comments:

Risk Management and its associated management programs and procedures are apart of the facility license and requirements thereby making this an SRO topic per 10 CFR55.43(b)(1).

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Radiation Control Group # N/A K/A # G2.3.6 Importance Rating 3.8 Ability to approve release permits.

Question #98 A radwaste release from the Discharge Monitor Tank 'A' is planned.

(1) Per HTP-ZZ-02006, Liquid Radwaste Release Permit (Batch), the release should be initiated within a MAXIMUM of ____(1)_____ after the release permit is generated.

And (2) If the release permit contains a "warning flag" for Total Body Dose, who is required to authorize the release?

A. (1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (2) Duty Manager B. (1) 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months (2) Radiation Protection Manager C. (1) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (2) Duty Manager D. (1) 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months (2) Radiation Protection Manager Answer: D Explanation:

Per HTP-ZZ-02006, step 4.2 "A Liquid RadWaste (Batch) should normally be initiated within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the release permit is generated. This time may be exceeded and the release permit considered valid upon verification that the tank contents have NOT been altered since the permit was generated." The distractor of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is from HTP-ZZ-02012, Gaseous radwaste permits, step 4.9 "Containment Purge or Vent release is initiated within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the release permit is generated" Per HTP-ZZ-02006, step 4.4 "RPM, or his designee, must authorize the release of any DMT with pre-release permit warning flags with Total Body Dose greater than 3E-4 mrem or Organ Dose

NRC Written Examination Callaway Plant Senior Reactor Operator greater 1E-3 mrem. Duty Manager is plausible as the shift manager is required to contact the duty manager is contact for various events and approvals on shift per the conduct of operations, ODP-ZZ-00001 Addendum 13.

A. Incorrect - See above explanation - both are wrong B. Incorrect - See above explanation - the time duration is wrong C. Incorrect - See above explanation - the approval is wrong D. Correct Technical Reference(s):

1. HTP-ZZ-02006, Liquid Radwaste Release Permit (Batch), Rev 94 References to be provided to applicants during examination: None Learning Objective: T61.003A, Normal Operations, LP #22, Objective A: DISCUSS the following concerning Liquid Waste Discharges:

1. Regulations and Callaway procedural controls governing:

a. Radiological Discharges

b. Chemical Discharges

1) National Pollution Discharge Elimination System (NPDES)

2) Comprehensive Environmental Response, Compensation and Liability Act (CERCLA)

c. Non-compliance consequences

2. Administration and Control

a. Liquid Radioactive Waste (LRW) Discharge Permit

b. NPDES Permit

3. Recent Liquid Waste Discharge Incidents

a. Root causes

b. Remedial actions

4. Precautions to prevent discharge violations Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____4____

10 CFR Part 55 Content:

NRC Written Examination Callaway Plant Senior Reactor Operator 10 CFR 55.43(b)(4)

Comments:

SRO Only per 10 CFR 55.43.4 - Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Reviewed Audit exam for overlap. Question #98 is focused on gaseous release via containment purge therefore no overlap concerns exist Operationally valid as operations initiates the release after the permit is generated and is responsible to ensure / verify the permit is valid (including administrative controls etc.) prior to the release

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Emergency Procedures/Plan Group # N/A K/A # G2.4.29 Importance Rating 4.4 Knowledge of the emergency plan.

Question #99 What is the LOWEST emergency classification at which the Shift Manager /

Emergency Coordinator MUST perform accountability of station personnel?

A. Unusual Event B. Alert C. Site Area Emergency D. General Emergency Answer: C Explanation:

Per EIP-ZZ-00230, Step 3.1. Shift Manager/Emergency Coordinator 3.1.1. Implements this procedure when any of the following implementing conditions are met:

For any Emergency Classification at the Shift Managers/Emergency Coordinators (SM/EC) discretion. [Ref: 5.2.2]

When there is a need to identify possible missing personnel.

When there is a need for Non-Essential Personnel to exit the Protected Area.

When there is a need for Non-Essential Personnel to exit the Exclusion Area.

When a SITE AREA EMERGENCY or GENERAL EMERGENCY is declared.

A. Incorrect - Plausible as several notifications of personnel (facility, control room communicator, onsite personnel and the duty manager) and the activation of ERDS per EIP-ZZ-00102.

B. Incorrect - Plausible as several ERO facilities (OSC, TSC, EOF) are staffed at the Alert level C. Correct D. Incorrect - Plausible as accountability is also performed at the GE level.

Technical Reference(s):

1. EIP-ZZ-00230, Accountability, Rev 35

2. EIP-ZZ-00102, Emergency Implementing Actions, Rev 65 References to be provided to applicants during examination: None

NRC Written Examination Callaway Plant Senior Reactor Operator Learning Objective: T68.1020, RERP, LP#1, Objective L: State when accountability is required, per EIPZZ00230.

Question Source: Bank # ___X _L22926__

Modified Bank # ______

New ______

Question History: Last NRC Exam ___N/A_________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____3___

10 CFR Part 55 Content:

Comments:

SRO Only due to conditions and limitations of the facility license. The emergency plan is part of the facility's license and the SRO's on shift may serve as the Emergency Coordinator during portions of the emergency and are response for crew accountability during the emergency

NRC Written Examination Callaway Plant Senior Reactor Operator Examination Outline Cross-reference: Level SRO Rev 0 Tier # 3 Emergency Procedures/Plan Group # N/A K/A # G2.4.23 Importance Rating 4.4 Knowledge of the bases for prioritizing emergency procedure implementation during emergency operations.

Question #100 Per ODP-ZZ-00025, EOP/OTO User's Guide, what is the reason a crew should NOT transition to the Functional Restoration Procedures (FRPs) from the Status Trees until E-0 Attachment A is complete?

A. No transition to FRPs is allowed because Status Trees are not monitored until the completion of E-0 Attachment A B. Verification of automatic actions of the safeguards equipment is the first means of preventing any challenges to the Critical Safety Functions (CSFs)

C. FRPs contain multiple redundant actions that are included in E-0 Attachment A and by completing E-0 Attachment A first these actions can be bypassed in the FRPs by the CRS D. To minimize the interface with other procedures thus providing the Control Room staff with an easily located set of instructions while minimizing the simultaneous use of other procedures Answer: B Explanation:

Per ODP-ZZ-00025, step 4.23.10,"If E-0 Attachment A, Automatic Action Verification is being performed and is not complete the user should not transition to the FRP from the Status Trees until Attachment A is complete. Verification of automatic actions ensures that plant equipment is operating properly. The proper operation of the safeguards equipment is the first means of preventing or correcting any challenges to the Critical Safety Functions."

The distractor of "No transition to FRPs is allowed because Status Trees are not monitored until the completion of E-0 Attachment A" is paraphrasing of step 4.24.8 which states:

Status Trees Monitoring

a. Monitoring shall begin as follows:

When directed in E-0, Reactor Trip or Safety Injection, to begin monitoring CSF Status Trees.

When E-0 is exited.

E-0 directs monitoring of status trees at Step #18 if neither of the above situation have occurred.

NRC Written Examination Callaway Plant Senior Reactor Operator The distractor of "FRPs contain multiple redundant actions that are included in E-0 Attachment A and by completing E-0 Attachment A first these actions can be bypassed in the FRPs by the CRS" is a rephrasing of Section 4.5, Redundant Actions in the EOPs, which states: "Redundant actions throughout the EOPs and multiple passes through the same path will cause the operator to encounter actions that have already been performed. These steps may be bypassed if the following conditions are met:

The CRS reads or paraphrases the step aloud and receives crew concurrence that the step has been performed.

The CRS does not bypass any checks for subsequent failures, such as Steam Generator pressure behavior checks."

The distractor of "To minimize the interface with other procedures thus providing the Control Room staff with an easily located set of instructions while minimizing the simultaneous use of other procedures" is from section 4.9, EOP Addendum, step #4.9.2 which states "The EOPs are designed to minimize the interface with other procedures. The EOPs include EOP Addenda that operators may need during recovery. This provides the Control Room staff with an easily located set of instructions while minimizing the simultaneous use of other procedures.

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

1. ODP-ZZ-00025, EOP/OTOs User Guide, Rev 37 References to be provided to applicants during examination: None Learning Objective: T61.003B, Off Normal Operations, LP #1, Objective E & F & G & H:

E. DESCRIBE the purpose and Scope of ODP-ZZ-00025, EOP/OTO Users Guide.

F. DISCUSS the responsibilities addressed in ODP-ZZ-00025, EOP/OTO Users Guide, for the following positions.

Shift Manager (SM)

Control Room Supervisor (CRS)

Incident Assessor (IA)

Reactor Operator (RO)

Operations Technician (OT)

G. DISCUSS the General Guidance section of ODP-ZZ-00025, EOP/OTO Users Guide, to include:

Indications

Plant parameters

Exiting OTOs

Tripping of bistables H. DISCUSS procedural requirements for ODP-ZZ-00025, EOP/OTO Users Guide, to include:

Command and Control

Step Sequencing

Immediate Actions

Placekeeping

Referencing and Branching

NRC Written Examination Callaway Plant Senior Reactor Operator

Transition to EOPs

NOTES and CAUTIONS

Blocking / Bypassing ESFAS

OTO Entry

Check / Ensure

Continuous Action Steps

A / ER and RNO Question Source: Bank # ______

Modified Bank # ______

New ___X____

Question History: Last NRC Exam ____N/A________

Question Cognitive Level:

Memory or Fundamental Knowledge __X___

Comprehension or Analysis _____

LOD ____2___

10 CFR Part 55 Content: