ML050490198

Wc 05-2004-SRO Final Written Reference Exam

ML050490198
Person / Time
Site:	Wolf Creek
Issue date:	05/13/2004
From:	NRC Region 4
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Download: ML050490198 (36)

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QUESTION # 76 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 1 K/A # 4.1 007/ 2.4.45 - Reactor Trip, Emergency Procedures / Plan: Ability to prioritize and interpret the significance of each annunciator or alarm.

Importance 3.6 Proposed Question:

The following plant conditions exist:

• Reactor has tripped. First Out Annunciator overtemperature differential temperature (OTT) has alarmed in Red.

• The crew has just transitioned to EMG ES-02, REACTOR TRIP RESPONSE from EMG E-0 REACTOR TRIP OR SAFETY INJECTION step 4.

• First Out Annunciator 088A, PZR PRESS SI RX TRIP has alarmed in white.

What actions should the Control Room Supervisor direct?

A. Transition back to EMG E-0, at Step 1.

B. Transition to EMG ES-03, SI TERMINATION, at Step 1.

C. Continue in EMG E-0, at Step 5.

D. Continue in EMG ES-02, at Step 1.

Proposed Answer: A, Transition back to EMG E-0, at Step 1.

Explanation: Answer A is correct. The CRS must recognize Annunciator 088A alarming would also indicate SI is actuating; the correct procedure response would be to re-enter EMG E-0 at step

1. Distracters B and C are possible choices for a CRS that recognizes SI has actuated but errors in applying procedure use. Distracter D is possible if a CRS believes that since 088A is white it is not important and would continue in the current procedure.

Technical

References:

ALR 00-088A Learning Objective: LO1732315, Objective 4 Question Source: NEW Question Cognitive Level: Comprehension or Analysis A3 10 CFR Part 55 Content: 55.43 (5)

Comments: SI is required if PZR level cannot be maintain above 6% or pressure cannot be maintained above 1830 psig.

QUESTION # 77 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 1 K/A # 4.1 009 EA 2.23 - Ability to determine or interpret the following as they apply to a small break LOCA: RCP operating parameters and limits.

Importance 3.3 Proposed Question:

The following conditions exist:

• Small break LOCA is in progress.

• Both CCPs are unavailable.

• B SI pump is OOS.

• RCS Subcooling is 30 °F.

The crew has just completed their transition brief from EMG E-0, REACTOR TRIP OR SAFETY INJECTION to EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT.

The Reactor Operator reports A SI pump has tripped and RCS pressure is 1400 psig and decreasing.

What action should be directed regarding the RCPs?

A. Leave all RCPs running since subcooling criteria is met.

B. Leave all RCPs running since no high head injection exists.

C. Stop all RCPs when RCS decreases below 1400 psig.

D. Stop all RCPs if RCS subcooling will decrease below limits.

Proposed Answer: B, Leave all RCPs running since no high head injection exists.

Explanation: Ans. C and D are incorrect because the foldout page criteria for tripping an RCP is not met for these conditions. Ans. A is incorrect since RCP trip criteria does apply in EMG E-1.

Ans. B is correct since without a running SI pump or any CCPs, high head injection is unavailable and RCPs must remain running.

Technical

References:

BD EMG E-1 page 72 & 73 and EMG E-1 foldout page Learning Objective: LO1732320, Objective 3 Question Source: NEW Question Cognitive Level: Comprehension or Analysis A2 10 CFR Part 55 Content: 55.43 (5)

QUESTION # 77 Comments: The SRO is expected to know and implement the items in the foldout page and background document. This action is required (not tripping the RCPs) by the foldout page and the incorrect answers are also explained in the background document.

QUESTION # 78 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 1 K/A # 4.1 029 EA2.09 - Ability to determine or interpret the following as they apply to a ATWS:

Occurrence of a main turbine/reactor trip.

Importance 4.4 Proposed Question:

The plant is operating at 43% power and increasing.

The Main Turbine now trips automatically due to a High Vibration condition.

A Reactor trip does NOT occur.

Which procedure will you direct?

A. EMG FR-S1, RESPONSE TO NUCLEAR POWER GENERATION/ATWT B. EMG E-0, REACTOR TRIP OR SAFETY INJECTION C. OFN MA-001, LOAD REJECTION OR TURBINE TRIP D. OFN AC-002, MAIN TURBINE HIGH VIBRATION Proposed Answer: C, OFN MA-001, LOAD REJECTION OR TURBINE TRIP Explanation: This question tests whether the candidate knows if an ATWS has occurred or not. In this case an ATWS has NOT occurred since the Automatic Reactor Trip signal is blocked below P-9 (50%), thereby making answer C correct. Answer A is incorrect, no ATWT exists. Answer B incorrect, since a reactor trip is not required. Answer D is incorrect, since this only applies if the main turbine is still running.

Technical

References:

OFN MA-001, step 1; Tech Spec Bases 3.3.2 #16 Learning Objective: LO1732411, Objective 2; SY1301200, Objective 4 Question Source: NEW Question Cognitive Level: Memory or Fundamental Knowledge K3 10 CFR Part 55 Content: 55.43 (5)

Comments: SRO needs to know what to direct if the reactor does not trip and whether it is expected or not.

QUESTION # 79 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 1 K/A # 4.1 038 EA2.16 - 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.

Importance 3.0 Proposed Question:

A large Steam Generator Tube Rupture has occurred on A S/G. The crew has transitioned through the EMG network and is in EMG E-3, STEAM GENERATOR TUBE RUPTURE. With the ruptured S/G indicating 99% Narrow Range level, the BOP reports that the A S/G Atmospheric Relief Valve (ARV) is beginning to open.

Per the guidance in EMG E-3, you would direct the BOP to ensure A S/G ARV controller is in:

A. MANUAL and closed.

B. MANUAL and full open.

C. AUTO and to increase the setpoint to 1185 psig.

D. AUTO and the setpoint is at 1125 psig.

Proposed Answer: D, AUTO and the setpoint is at 1125 psig.

Explanation: Ans. D is correct. The procedure verifies the ARV is available in Auto to protect the S/G Safeties. A stuck open safety would result in a continuous uncontrolled release. There is some misconceptions exhibited by operators in the desire to terminate the release if an ARV begins opening. In a SGTR, closing the ARV would then cause a challenge to the safeties.

Technical

References:

EMG E-3 step 3, BD EMG E-3 step 3 Learning Objective: LO1732325, Objective 5 Question Source: NEW Question Cognitive Level: Memory or Fundamental Knowledge K3 10 CFR Part 55 Content: 55.43 (5)

Comments: None

QUESTION # 80 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 1 K/A # 4.2 056 / 2.1.32 - Loss of Offsite Power, Conduct of Operations: Ability to explain system limits and precautions.

Importance 3.8 Proposed Question:

The Unit is at 50% power with the 69kV line down for maintenance. A severe thunderstorm west of the plant causes a loss of both the Benton and Rose Hill 345 kV lines.

What is the status of the Offsite Power Sources with regards to Technical Specifications (T.S.)?

A. One offsite circuit inoperable requiring entry into T.S. 3.8.1.

B. Two Offsite sources are lost requiring entry into T.S. 3.0.3.

C. A Potential Offsite source is lost, enter T.S. 3.0.6 and perform a Safety Function Determination.

D. Two Offsite sources are still available and no T.S. action is required.

Proposed Answer: D, Two Offsite sources are still available and no action is required.

Explanation: Answer D is correct per the T.S. BASES for operable offsite sources. The offsite source goes from the switchyard transformers to the vital AC bus. Both transformers are still energized given the starting conditions. Answers A and B are incorrect if the CRS believes there has been a loss of offsite power. Answer C is incorrect but a plausible distracter for the CRS that does not fully understand Safety Function Determinations.

Technical

References:

T.S. 3.8.1 and Bases Learning Objective: SY1506201, Objective 7 Question Source: Modified Bank Q19109 Question Cognitive Level: Memory or Fundamental Knowledge K3 10 CFR Part 55 Content: 55.43 (2)

Comments: None

QUESTION # 81 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 1 K/A # 4.2 057 AA2.18 - Ability to determine and interpret the following as they apply to the Loss of Vital AC Instrument Bus: The indicator, valve, breaker, or damper position which will occur on a loss of power.

Importance 3.1 Proposed Question:

The plant is at 100% with the Pressurizer level and pressure selected to the following positions:

• PZR LEV CTRL SEL, BB LS-459D, selected to L459/L460

• PZR PRESS CTRL SEL, BB PS-455F, selected to P455/P456 An event occurs and the control room staff notes the following indications:

• Charging pump suction has swapped to the RWST.

• Letdown flow remains stable for current plant conditions.

• Annunciator 00-076A, SSPS B GENERAL WARNING is in alarm.

• All S/G levels are stable for current plant conditions.

Based on these conditions which procedure must be implemented and why?

A. OFN NK-020, due to a loss of 125 VDC Bus NK02 B. OFN NN-021, due to a loss of 120 VAC Instrument Bus NN02 C. OFN NN-021, due to a loss of 120 VAC Instrument Bus NN04 D. OFN PK-029, due to a loss of 120 VAC Instrument Bus PN02 Proposed Answer: C, OFN NN-021, due to a loss of 120 VAC Instrument Bus NN04 Explanation: Since none of the instruments are selected to NN04, the only indication will be the annunciator and charging pump suction swapping making answer C the only correct answer.

Answers A, B and D are incorrect, but plausible if the candidate is not aware of which 120VAC feeds which indicator as selected on the select switches.

Technical

References:

OFN NN-021, Attachment A, page 1 Learning Objective: LO1732431, Objective 3 Question Source: NEW Question Cognitive Level: Comprehension or Analysis A2 10 CFR Part 55 Content: 55.43 (5)

QUESTION # 81 Comments: OFN NN-021 gives directions on what to do following a loss of NN04, but the SRO is still required to know what components will be affected.

QUESTION # 82 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 1 K/A # 4.5 E04 (LOCA Outside Containment)/ 2.4.10 - Emergency Procedures/Plan: Knowledge of annunciator response procedures.

Importance 3.1 Proposed Question:

Given the following plant conditions:

• A Safety Injection has occurred.

• EMG E-0, REACTOR TRIP OR SAFETY INJECTION is in progress.

• CTMT Radiation Alarms are clear.

• PRT Alarms are clear.

• CTMT Sump Alarms are clear.

• SD RE-10, Aux Bldg Radiation Monitor is in alarm.

• Annunciator 00-096A, RHR RM SUMP A/B LEV HI is in alarm.

• RCS Pressure is 1800 psig and slowly decreasing.

• Subcooling is 85 degrees.

• PZR Level is 25% and slowly decreasing.

Based on the above indications, what mitigation path will be used?

A. EMG C-12, LOCA OUTSIDE CONTAINMENT B. EMG E-1, LOSS OF REACTOR OR SECONDARY COOLANT C. EMG ES-03, SI TERMINATION D. Continue in EMG E-0, and refer to OFN BB-007, RCS LEAKAGE HIGH Proposed Answer: A, EMG C-12, LOCA OUTSIDE CONTAINMENT Explanation: Answer A is correct since this is an indication of a leak outside containment. Ans. B is incorrect, this procedure a possible transition but would be entered if the leak were inside containment. Ans. C is incorrect, but plausible, since RCS conditions are met for transition, however other indications are available that a LOCA is still in progress. Ans. D is incorrect, but plausible since an OFN can be used with an EMG procedure, but in this case the indications are clear that a break has occurred outside containment.

Technical

References:

EMG E-0, Step 25 Learning Objective: LO1732333, Objective 2 Question Source: Bank #Q20324 Question Cognitive Level: Comprehension or Analysis A3

QUESTION # 82 10 CFR Part 55 Content: 55.43 (5)

Comments: The SRO must be able to use control room alarms to assist in diagnosing the event.

This question also examines the SRO knowledge of mitigation flow paths in the Emergency procedures.

QUESTION # 83 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 2 K/A # 4.2 024 AA2.05 - Ability to determine and interpret the following as they apply to the Emergency Boration: Amount of boron to add to achieve required SDM Importance 3.9 Proposed Question:

The following plant conditions exist:

• Unit is in Mode 6.

• Emergency boration has been started due to a low boron concentration in the Refueling Pool.

• Boration flow has been verified.

Using the guidance contained in OFN BG-009, EMERGENCY BORATION, what condition must be established in order for you to verify Shutdown Margin is adequate?

A. Start up rate must be decreasing.

B. Shutdown margin must be verified to be greater than 1.3%?K/K.

C. Keff must be verified to be less than 0.99.

D. RCS boron concentration must be greater than 2300 ppm.

Proposed Answer: D, RCS boron concentration must be greater than 2300 ppm.

Explanation: Answer A is incorrect as Startup rate can be zero OR decreasing to stop boration flow. Answer B is correct for all conditions except mode 6. Answer C is incorrect as Keff must be less than 0.95 to stop flow. Answer D is correct per procedure.

Technical

References:

OFN BG-009 entry conditions and step 10 Learning Objective: LO1732419, Objectives 2 & 5 Question Source: NEW Question Cognitive Level: Comprehension or Analysis A2 10 CFR Part 55 Content: 55.43 (7)

Comments: The SRO must remember that the criteria changes due to changing plant conditions and the correct criteria applied for the existing conditions.

QUESTION # 84 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 2 K/A # 4.2 028 AA2.12 - Ability to determine and interpret the following as they apply to the Pressurizer Level Control Malfunctions: Cause for PZR level deviation alarm: controller mal-function or other instrumentation malfunction Importance 3.5 Proposed Question:

The unit is stable at 100% power when you receive alarm 00-032C, PZR LO LVL DEV.

The RO reports the following indications exist:

• LI-459A 58%

• LI-460A 59%

• LI-461A 51%

Based on these indications, determine the problem and the proper procedure to mitigate the event?

A. The non-controlling PZR level channel is failing LOW, enter ALR 00-032C.

B. The controlling PZR level channel is failing LOW, enter OFN SB-008, INSTRUMENT MALFUNCTIONS.

C. Charging Flow Control Valve BG FCV-121 is failing OPEN, enter OFN SB-008, INSTRUMENT MALFUNCTIONS.

D. Charging Flow Control Valve BG FCV-121 is failing CLOSED, enter ALR 00-032C.

Proposed Answer: B, The controlling PZR level channel is failing LOW, enter OFN SB-008, INSTRUMENT MALFUNCTIONS.

Explanation: Alarm 00-032C is driven from the controlling channel deviating by more than 5%

from program. Answer A is incorrect since it is the non-controlling channel. Answer B is correct.

Answer C & D could cause the deviation alarm, but all three channels would read the same.

Technical

References:

ALR 00-032C, OFN SB-008 Learning Objective: LO1732418, Objective 3 Question Source: NEW Question Cognitive Level: Comprehension or Analysis A3 10 CFR Part 55 Content: 55.43 (5)

Comments: Operators must be able to differentiate between a system failure and a controller failure based on control room indications.

QUESTION # 84 QUESTION # 85 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 2 K/A # 4.2 051 / 2.1.32 - Loss of Condenser Vacuum / Conduct of Operations: Ability to explain and apply all system limits and precautions.

Importance 3.8 Proposed Question:

The plant is operating at 80% power when a failure of the condenser boot seal causes a loss of vacuum and subsequent turbine trip.

What is the Technical Specification bases for the reactor trip and which signal directly causes the reactor to trip?

A. Prevention of a pressure and temperature transient on the reactor because of a 2 out of 4 Turbine Control Valves closed signal.

B. Prevention of a pressure and temperature transient on the reactor because of a 2 out of 4 Turbine Stop Valves closed signal.

C. Anticipation of a loss of RCS heat removal because of a 2 out of 3 EHC fluid oil pressure low signal.

D. Anticipation of a loss of RCS heat removal because of a main condenser vacuum low signal.

Proposed Answer: C, Anticipation of a loss of RCS heat removal because of a 2 out of 3 EHC fluid oil pressure low signal.

Explanation: Answers A, B and D are incorrect because none of these directly trips the reactor.

The bases, as seen in Tech Specs, identifies that the trip is in anticipation of loss of heat sink vs.

an excessive cooldown as one might expect. Therefore the only correct answer is C per Tech spec. 3.3.1 and its bases.

Technical

References:

Tech Spec 3.3.1 and associated bases Learning Objective: LO1732204, Objective 1 Question Source: NEW Question Cognitive Level: Comprehension or Analysis A3 10 CFR Part 55 Content: 55.43 (2)

Comments: Power level given at > 50% power to ensure that P-9 does not prevent Reactor trip.

Bases and actual trip knowledge

QUESTION # 86 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 2 K/A # 4.2 076 / 2.4.9 - High Reactor Coolant Activity / Emergency Procedures / Plan:

Knowledge of low power / shutdown implications in accident (e.g. LOCA or loss of RHR) mitigation strategies.

Importance 3.9 Proposed Question:

A rapid plant shutdown was required due to the indication of a loose part in the Reactor Coolant system.

The plant is currently in Mode 3 with RCS temperature at 525 °F and the operators are cooling down to enable Steam Generator opening to find the loose part.

Chemistry has taken a sample and informed you that RCS Activity is currently reading 30 µCi/gm greater than the 100/ E-Bar value calculated earlier.

Which of the following describes the applicable Emergency Plan Classification required for the current plant conditions?

A. Alert B. Notification of Unusual Event C. Not applicable, due to the plant in Mode 3 D. Not applicable, due to reactor power less than 20%

Proposed Answer: B, Notification of Unusual Event.

Explanation: High activity greater than a Tech Spec limit requires a NOUE classification regardless of Mode or power level. Answers C & D may seem plausible since the graph in Tech specs only reads down to 20% power and the Tech spec is not applicable < than 500 degrees F.

Answer A is incorrect, since the Alert level would require another significant event in order to get to this level which is not indicated in the stem.

Technical

References:

Tech Spec 3.4.16 and EAL-5 of APF 06-002-01 Learning Objective: LR1007001, Objective 1 Question Source: NEW Question Cognitive Level: Comprehension or Analysis A3 10 CFR Part 55 Content: 55.43 (5)

QUESTION # 86 Comments: Candidate needs to know when an Emergency Classification is required to performed.

QUESTION # 87 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 1 Group # 2 K/A # 4.5 W/E15 2.3.4 - Containment Flooding / Radiation Control: Knowledge of radiation exposure limits and contamination control, including permissible levels in excess of those authorized.

Importance 3.1 Proposed Question:

The following conditions exist:

A Large Break LOCA has occurred inside containment. An operator has been sent into the Aux Building Penetration room to close a valve on a system that is causing flooding in containment and manually isolate a valve that will prevent a massive offsite dose to the general population.

The operator injures himself such that his life is in jeopardy. Another operator has volunteered to extricate the individual from the area. This volunteer operator has a current year to date exposure of 3 REM.

How much dose can you as the Emergency Manager permit this volunteer to receive?

A. 7 REM B. 10 REM C. 22 REM D. 25 REM Proposed Answer: D, 25 REM Explanation: A volunteer is allowed to receive a single dose of 25 REM for lifesaving, which is answer D. Answer A is for saving equipment minus 3 REM. Answer B is for saving valuable equipment. Answer C is the max dose given minus the 3 REM. Answer D is correct even though a higher dose may be authorized it is still the highest dose given.

Technical

References:

AP 06-002, step 6.3.15 and attachment E Learning Objective: GE1135628, Objective 2 Question Source: NEW Question Cognitive Level: Memory or Fundamental Knowledge K3 10 CFR Part 55 Content: 55.43 (4)

Comments: Knowledge of Emergency Limits and how they are implemented.

QUESTION # 88 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 2 Group # 1 K/A # 3.4 003 A2.03 - Ability to (a) predict the impacts of the following malfunctions or operations on the RCPS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: Problems associated with RCP motors, including faulty motors and current, and winding and bearing temperature problems.

Importance 3.1 Proposed Question:

Unit load is currently 650 Mwe.

A Reactor Coolant Pump (RCP) has the following conditions:

• Motor Upper Radial Bearing 190°F

• Motor Upper Thrust Bearing 191°F

• Motor Lower Radial Bearing 196°F

• Motor Lower Thrust Bearing 192°F

• Motor Stator Winding Temperature 195°F

• Number 1 Seal and Bearing Water Temperature 198°F The crew enters OFN BB-005, RCP MALFUNCTIONS.

What is the proper action for these conditions?

A. Commence plant shutdown using OFN MA-038, RAPID PLANT SHUTDOWN, when less than 48% trip the A RCP.

B. Trip reactor, enter EMG E-0, REACTOR TRIP OR SAFETY INJECTION, then trip A RCP while concurrently using OFN BB-005.

C. Continue to monitor RCP bearings and windings in accordance with OFN BB-005 and trip the pump if any temperature exceeds 200°F.

D. Trip A RCP in accordance with OFN BB-005, then trip the reactor and enter EMG E-0, REACTOR TRIP OR SAFETY INJECTION.

Proposed Answer: B, Trip reactor, enter EMG E-0, REACTOR TRIP OR SAFETY INJECTION, then trip A RCP while concurrently using OFN BB-005.

Explanation: Ans. A and C incorrect, since a reactor trip, not a controlled shutdown, is required for this condition. Ans D incorrect, because the Reactor must be tripped prior to tripping an RCP.

B is correct since any motor bearing temperature > 195°F is RCP trip criteria, but the Reactor must be tripped first.

Technical

References:

OFN BB-005, RCP MALFUNCTIONS Learning Objective: LO1732415, Objective 2

QUESTION # 88 Question Source: Bank #22749 (modified)

Question Cognitive Level: Comprehension or Analysis A3 10 CFR Part 55 Content: 55.43 (5)

Comments: SROs need to know RCP trip criteria when they see it.

QUESTION #89 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 2 Group # 1 K/A # 3.2 004 A2.09 - Ability to (a) predict the impacts of the following malfunctions or operations on the CVCS; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those malfunctions or operations: High primary and/or secondary activity.

Importance 3.9 Proposed Question:

A rapid (5%/min.) power reduction from 100% to 60% was performed due to grid instabilities.

Power has been stable at 60% for seven hours.

Calculated 100/E-bar value is 250 µCi/gm.

Chemistry reports indications of higher RCS activity.

The results from the RCS chemistry samples, taken four hours after power was stabilized at 60%,

are as follows:

• Dose equivalent I-131 (DEI) 97 µCi/gm

• Gross coolant activity 45 µCi/gm

• Mixed Bed Demineralizer - Cesium Decontamination Factor 15

• Mixed Bed Demineralizer - Iodine Decontamination Factor 18 What are the required actions?

(Figure 3.4.16 of Tech Specs is provided)

A. The DEI limit has been exceeded, shutdown the plant in accordance with Tech Specs using GEN 00-004, POWER OPERATION.

B. The 100/E-Bar gross activity limit has been exceeded, shutdown the plant in accordance with Tech Specs using GEN 00-004, POWER OPERATION.

C. The 100/E-Bar gross activity limit has been exceeded, maximize cleanup flow as directed by Chemistry in accordance with OFN BB-006, HIGH REACTOR COOLANT ACTIVITY.

D. Mixed Bed Demineralizer Decontamination Factors are out of specification, shift Demineralizers in accordance with OFN BB-006, HIGH REACTOR COOLANT ACTIVITY.

Proposed Answer: C, The 100/E-Bar gross activity limit has been exceeded, maximize cleanup flow as directed by Chemistry in accordance with OFN BB-006 HIGH REACTOR COOLANT ACTIVITY.

Explanation:. Ans. A is incorrect since DEI has not been exceeded. Ans B is incorrect, since the 45 µCi/gm is less than the given 100/E-bar given. Answer C is correct since activity is greater than 10% of the 100/E-bar value given. Ans. D is incorrect since the Decontamination factors are well above the values required to shift beds.

Technical

References:

OFN BB-006, Tech Spec. 3.4.16 and Figure 3.4.16-1(provided)

QUESTION #89 Learning Objective: LO1733203, Objective 3 Question Source: Bank #Q24390 (modified)

Question Cognitive Level: Comprehension or Analysis A3 10 CFR Part 55 Content: 55.43 (5)

Comments: Predict the upcoming actions contained within abnormal operating procedures given a large decrease in power and expected high RCS activity.

QUESTION #90 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 2 Group # 1 K/A # 3.2 013 A2.06 - 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: Inadvertant ESFAS actuation.

Importance 4.5 Proposed Question:

I&C technicians are performing surveillance testing on the Main Steam line pressure transmitters and have inadvertently de-energized 2 of 3 transmitters on Main Steam line C. The reactor tripped and the crew has entered EMG E-0, REACTOR TRIP OR SAFETY INJECTION. The first four steps of EMG E-0 have been completed and the following conditions are observed:

• RCS pressure is 2000 psig and trending up

• S/G pressures are 1000 psig and trending up

• Containment pressure is normal

• RCS subcooling is 70°F and increasing slowly

• Pressurizer level is 35% and trending up What mitigation path should be followed?

A. Shut the BIT inlet valves and transition to EMG ES-02, REACTOR TRIP RESPONSE, since Safety Injection is not required.

B. Leave BIT inlet valves open and continue in EMG E-0 since a Safety Injection is required.

C. Shut the BIT inlet valves and continue in EMG E-0 since a Safety Injection is not required.

D. Shut the BIT inlet valves and transition to EMG ES-03, SI TERMINATION, since Safety Injection is not required.

Proposed Answer: C, Shut the BIT inlet valves and continue in EMG E-0 since a Safety Injection is not required.

Explanation: The candidate must realize that not only did a reactor trip occur, but that a safety injection was actuated inadvertently. The BIT valves should be closed following this inadvertent SI to prevent overfilling the PZR via the CCPs. SI must be terminated quickly and is addressed at step 5 of EMG E-0. Answer C is then correct. Answer B is incorrect since the BIT inlet should be closed. Answer A and D are incorrect since EMG E-0 will address the problem.

Technical

References:

EMG E-0, step1-5 and BD-EMG E-0, page 28 Learning Objective: LO1732313, Objective 3 Question Source: NEW Question Cognitive Level: Comprehension or Analysis A2

QUESTION #90 10 CFR Part 55 Content: 55.43 (5)

Comments: Need to analyze plant conditions and use appropriate procedures to mitigate.

QUESTION # 91 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 2 Group # 1 K/A # 073 2.3.9 - Process Radiation Monitoring System / Radiation Control: Knowledge of the requirements for reviewing and approving release permits.

Importance 3.1 Proposed Question:

Who is responsible for approving a radioactive release if all conditions for the release are satisfactory?

A. On shift Chemistry Technician B. Chemistry Supervisor C. Control Room Supervisor D. Shift Manager Proposed Answer: D, Shift Manager Explanation: Answer A is incorrect as this individual prepares the release. Answer B is incorrect as this individual approves a release if the checklist items are not all met. Answer C is incorrect as this individual is responsible for performing the release and ensuring all data is correctly entered.

Answer D is correct.

Technical

References:

AP 07B-001, Radioactive Releases, Section 5.6 Learning Objective: LO1733204, Objective 10 Question Source: NEW Question Cognitive Level: Memory or Fundamental Knowledge K2 10 CFR Part 55 Content: CFR 55.43 (4)

Comments: None

QUESTION #92 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 2 Group # 2 K/A # 3.5 028 / 2.3.9 - Recombiner and Purge Control / Radiation Control: Knowledge of the process for a performing a containment purge.

Importance 3.5 Proposed Question:

A Containment Purge is in progress to reduce containment pressure per SYS GT-120, CONTAINMENT MINI PURGE OPERATIONS.

Which of the following conditions would require you to direct Chemistry to sample the containment atmosphere?

A. Containment Purge stopped momentarily then was re-started.

B. Reactor power change of greater than 15% in one hour.

C. Plant mode change from Mode 3 to Mode 4.

D. Purge supply air temperature decreases to less than 50°F.

Proposed Answer: B, Reactor power change of greater than 15% in one hour.

Explanation: Ans. A is incorrect, since a purge is allowed to be started and stopped without having to sample. Ans. B is correct per AP 07B-001, RADIOACTIVE RELEASES. Ans. C is incorrect since the Mode changes that require sampling are changes from Mode 3-2 or Mode 2-3. Ans. D is incorrect, since the temperature at which the mini purge supply fan trips is 40°F not the 50°F given which would not require a sample.

Technical

References:

AP-07B-001, step 6.2.4, #5 Learning Objective: SY1302800, Objective 7 Question Source: NEW Question Cognitive Level: Comprehension or Analysis A2 10 CFR Part 55 Content: 55.43 (4)

Comments: None

QUESTION # 93 Question Worksheet Examination Outline Cross-reference: Level RO Tier # 2 Group # 2 K/A # 3.4 055 / 2.2.25 - Condenser Air Removal / Equipment Control: Knowledge of bases in technical specifications for limiting conditions for operations and safety limits.

Importance 2.5 Proposed Question:

Which of the following conditions must be met in order for the Condenser Air Discharge Radiation Monitor, GE RE-92, to be operable?

It must be able to:

A. detect S/G leakrates of 30 gpd at existing RCS activity levels.

B. alarm if the S/G leakrate increases by 5 gpd.

C. detect a S/G leakrate of 1 gpm within 15 minutes of the leak initiation.

D. isolate S/G blowdown upon receipt of a high radiation alarm.

Proposed Answer: A, detect S/G leakrates of 30 gpd at existing RCS activity levels.

Explanation: Answer A is the correct answer per TRM. Answer B is incorrect as the leakrate detection is tied to 30 gpd, not 5 gpd. Answer C is incorrect as no time limit is mentioned for detection. Answer D is incorrect and is a true statement, but the isolation of blowdown function is not required for operability Technical

References:

TRM 3.3.18 bases Learning Objective: SY1505500, Objective 5 Question Source: NEW Question Cognitive Level: Memory or Fundamental Knowledge K3 10 CFR Part 55 Content: 55.43 (2)

Comments: Operability of secondary rad monitors is needed to identify S/G primary to secondary leaks. These leaks are known to escalate rapidly if quick action is not taken. Taking action is based on knowing that a leak exists.

QUESTION # 94 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 3 Group # 1 K/A # 2.1.12 - Conduct of Operations: Ability to apply technical specifications for a system.

Importance 4.0 Proposed Question:

Given the following:

• The plant is at 100% power.

• NIS Power Range Channel N-43 experienced a failed power supply and has been removed from service for corrective maintenance.

• NIS Power Range Channel N-44 power indication has started oscillating between 80% and 100%. The STA confirms the same oscillation at the N-44 panel.

Which one of the following actions is required?

A. Trip the reactor and enter EMG E-0, REACTOR TRIP OR SAFETY INJECTION.

B. Bypass N-43 for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while troubleshooting N-44. If the channel cannot be repaired within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Enter T.S. 3.0.3 and be in HOT STANDBY within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

C. Bypass N-44 for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while repairing N-43. If the channel cannot be repaired within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, Enter T.S. 3.0.3 and be in HOT STANDBY within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

D. Enter T.S. 3.0.3. Repair EITHER channel or be in HOT STANDBY within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

Proposed Answer: D, Enter T.S. 3.0.3. Repair EITHER channel or be in HOT STANDBY within 7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />.

Explanation: A. Incorrect - No trip signal is present. B & C Incorrect - The note in T.S. 3.3.1 allows bypassing a failed channel for routine surveillance testing, not troubleshooting or repairs. D.

Correct - There are no Actions for two channels inoperable and would require entry into T.S.

3.0.3.

Technical

References:

T.S. 3.3.1, Reactor Trip System Instrumentation, and Bases.

T.S. 3.0.3 and Bases Learning Objective: SY1301501, Objective 13 Question Source: INPO Bank - Salem 11/04/02 Question Cognitive Level: Comprehensive or Analysis A3 10 CFR Part 55 Content: 55.43 (2)

Comments: This question tests the SROs ability to recognize plant conditions that are outside the actions allowed in Technical Specifications requiring, entry into T.S. 3.0.3.

QUESTION # 94 QUESTION # 95 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 3 Group # 1 K/A # 2.1.14 - Conduct of Operations: Knowledge of system status criteria which require the notification of plant personnel.

Importance 3.3 Proposed Question:

Under which of the following conditions would a Plant Announcement NOT be made?

A. Declaration of a Notification of Unusual Event (NUE) due to a Chemical Spill located outside the Protected Area but upwind.

B. Declaration of a Site Area Emergency (SAE) during inclement winter weather.

C. Declaration of a Site Area Emergency (SAE) during a declared Security Emergency.

D. Declaration of a Notification of Unusual Event (NUE) due to a Tech Spec required plant shutdown.

Proposed Answer: C, Declaration of a Site Area Emergency (SAE) during a declared Security Emergency.

Explanation: All declared emergencies are announced to the plant in order to activate resources for the Control Room. The exception is during a Security Emergency where intruders have entered the protected area. Announcing the declaration could cause confusion for emergency responders to attempt to get to their assigned positions. Answer B is testing a misconception about severe weather. During inclement weather it may be better NOT to make a Protective Action Recommendation (PAR) to evacuate, the emergency still has to be declared and announced to activate the TSC on site.

Technical

References:

OFN SK-039 Learning Objective: LO1732447, Objective 7 Question Source: NEW Question Cognitive Level: Memory or Fundamental Knowledge K3 10 CFR Part 55 Content: 55.43 (5)

Comments: None

QUESTION # 96 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 3 Group # 2 K/A # 2.2.18 - Equipment Control: Knowledge of the process for managing maintenance activities during shutdown operations.

Importance 3.6 Proposed Question:

The unit is currently at Mid-loop conditions with S/G Nozzle Dam installation in progress at the start of a refueling outage. Engineering has requested permission to start a Local Leak Rate Test (LLRT) on containment penetration P-32, Containment Sump Discharge.

The LLRT:

A. may not be performed until the unit has exited Mid-loop conditions.

B. may not be performed until the nozzle dams are installed.

C. may be performed if an individual is assigned to isolate the penetration within 30 minutes.

D. may be performed without additional monitoring since it is not a safety related system.

Proposed Answer: C, may be performed if an individual is assigned to isolate the penetration within 30 minutes.

Explanation: Containment Closure must be attainable within 30 minutes whenever the unit is in mid-loop conditions for all penetrations making Answer C correct.. Answers A and B are incorrect since the work is allowed. Answer D is incorrect since it makes no difference whether the system is safety related or not.

Technical

References:

GEN 00-008, step 4.4.3.; STS GP-006, step 6.2 Learning Objective: LO1732108, Objective 3 Question Source: NEW Question Cognitive Level: Memory or Fundamental Knowledge K3 10 CFR Part 55 Content: 55.43 (5)

Comments: Wolf Creek has made commitments due to previous problems in maintaining containment closure during refuelings.

QUESTION # 97 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 3 Group # 2 K/A # 2.2.33 - Equipment Control: Knowledge of control rod programming.

Importance 2.9 Proposed Question:

Which ONE of the following describes why the bank overlap unit withdraws control rod banks sequentially?

A. Provides input for control rod insertion limit alarms and control bank deviation alarms.

B. Provides for reduced oscillation in the size and location of peak power production and an input for control bank deviation alarms.

C. Provides for uniform rod worth and provides adequate Shutdown Margin.

D. Provides for uniform rod worth and maintains acceptable peak power production during rod motion.

Proposed Answer: D, Provides for uniform rod worth and reduced oscillation in the size and location of peak power production.

Explanation: Answers A and B are incorrect due to the fact the pulse to analog converter provides inputs to the Rod Insertion Limits and bank deviation alarms. Answer C is incorrect as Rod Insertion Limits ensure adequate Shutdown Margin Technical

References:

T.S. Bases 3.1.6 LCO Learning Objective: SY1300100, Objective 2 Question Source: INPO Bank - Summer 9/17/02 Question Cognitive Level: Memory or Fundamental Knowledge K3 10 CFR Part 55 Content: 55.43 (6)

Comments: Rod Control Programming

QUESTION # 98 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 3 Group # 1 K/A # 2.3.1 - Knowledge of 10 CFR: 20 and related facility radiation control requirements.

Importance 3.0 Proposed Question:

A Reactor Operator has been on loan to Callaway plant moving fuel in their refueling outage. He has been called back to Wolf Creek because of an emergent situation. Callaway will send his dose record for the visit within a week. You now need him to perform a task with a projected total dose of 700 mr.

Can you direct this individual to perform this task?

A. No, because their radworker quals are no longer valid.

B. No, because the total dose he can receive is 500 mr.

C. Yes, because 1000 mr is allowed for emergent work.

D. Yes, because the projected dose is less than 2000 mr.

Proposed Answer: B, No, because the total dose he can receive is 500 mr.

Explanation: Answer A is incorrect because his quals are still valid. Answer C and D are not correct because the 10 CFR limit is 500 mr for radworker that dont have full documentation.

Technical

References:

AP 25A-001 Learning Objective: LO1733204, Objective 4 Question Source: NEW Question Cognitive Level: Memory or Fundamental Knowledge K3 10 CFR Part 55 Content: 55.43 (4)

Comments: Need to know an individuals allowed dose after returning from another plant.

QUESTION # 99 Question Worksheet Examination Outline Cross-reference: Level SRO Tier # 3 Group # 3 K/A # 2.3.8 - Radiation Control: Knowledge of the process for performing a planned gaseous radioactive release.

Importance 3.2 Proposed Question:

You are directing the performance of SYS HA-204, GASEOUS RADWASTE SYSTEM DECAY TANK RELEASE PROCEDURE in preparations to release the contents of gaseous decay tank TH01B (#2)

On Monday the following occurs:

• At 1400 - the decay tank sample is completed and you authorize the Radioactive Gas Release Permit.

• At 1600 - the valve lineup to release the tank has been completed, but not started.

• Due to equipment malfunction the release is delayed.

On Wednesday at 1700 the Radwaste Operator requests permission to start this release.

Why wont you allow the release to occur?

A. Another gaseous decay tank sample must be taken.

B. The Radioactive Gas Release Permit has expired.

C. A valve lineup to release decay tank TH01B must be re-performed.

D. Radwaste Building Exhaust Monitor, GH RE-10B, must be re-calibrated.

Proposed Answer: A, Another gaseous decay tank sample must be taken.

Explanation: Answer A is correct, since if the release has not begun within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, another sample is required. Answer B is incorrect since the permit does not expire. Answer C is incorrect, since the procedure completed 2 days earlier is still good. Answer D is incorrect, since there is no reason to believe GHRE10B is out of calibration.

Technical

References:

AP 07B-001 step 6.2.3 Learning Objective: LO1733204, Objective 10 Question Source: NEW Question Cognitive Level: Comprehension or Analysis A2 10 CFR Part 55 Content: 55.43 (4)

QUESTION # 99 Comments: SRO personnel are responsible for authorizing the start and end of releases from gaseous sources and must know the limits associated with such releases.

QUESTION # 100 Question Worksheet Examination Outline Cross-reference: Level RO Tier # 3 Group # 4 K/A # 2.4.30 - Emergency Procedures / Plan: Knowledge of which events related to system operations/status should be reported to outside agencies.

Importance 3.6 Proposed Question:

Given the following conditions:

• Wolf Creek is at 100% power.

• Air in-leakage to the condenser has resulted in steadily degrading condenser vacuum.

• A load reduction is directed in order to maintain vacuum.

• With the unit at approximately 85% power, a manual reactor trip is ordered due to the inability to maintain vacuum.

• All systems function as designed.

Based solely on the information given, which of the following describes the notification requirements for this event?

A. No notifications to any outside agencies are required for these conditions.

B. The NRC must be notified within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> due to manual actuation of the Reactor Protection System.

C. System Operations - Topeka must be notified within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> in order to ensure grid stability.

D. The State / County must be notified within 15 minutes of the trip due to reaching an Emergency Plan classification for an Unusual Event.

Proposed Answer: B, The NRC must be notified within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> due to manual actuation of the Reactor Protection System.

Explanation: Per AP 26A-001,REPORTABLE EVENTS - EVALUATION AND DOCUMENTATION Attachment A, a four hour notification is required making answer B correct. Answer A is incorrect since notification is required. Answer C is a good idea, but not required. Answer D is incorrect since these conditions will not constitute a Notification of Unusual Event.

Technical

References:

AP 26A-001, Attachment A Learning Objective: LO1733214, Objective 2 Question Source: INPO Bank - Point Beach 09/29/03 Question Cognitive Level: Comprehension or Analysis A2 10 CFR Part 55 Content: 55.43 (5)

QUESTION # 100 Comments: SROs are required to know proper offsite notifications for given conditions.