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2015 Duane Arnold Energy Center Initial License Examination - Administered Exam
	ML15125A418
Person / Time
Site:	Duane Arnold
Issue date:	04/14/2015
From:	; Division of Reactor Safety III
To:	; NextEra Energy Duane Arnold
	Zoia C
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ML15118A356	List: ML15125A418; ML15232A162; ML15267A145; ML15267A167;
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WRITTEN / ORAL EXAMINATION KEY Page 1 COVER SHEET Examination Number/Title: PDA 15-1 RO NRC Written Exam Training Program: Licensed Operator Initial Training Course/Lesson Plan Number(s): 50007 Reactor Operator Total Points Possible: 75 PASS CRITERIA: 80%. Exam Time: 360 minutes Yes No Yes No This is an alternate examination; verified This is a remediation exam; verified at at least 30% of the questions are least 90% of the questions are different different from other forms/versions of this from the failed exam. For LOIT remedial exam (e.g., Forms A, B, C; continuing exams, verified 95% difference. For training exam versions for consecutive LOCT annual operating and biennial weeks). For LOCT weekly exams during comprehensive remedial exams, verified a segment, verified > 50% difference. no repeat questions.

This is an initial training examination; This is a randomly generated electronic verified at least 30% of the questions are exam printout; verified the exam bank has different from same exam administered 3 questions per objective if one test item to the previous class. on exam for the objective. If 2 or more test items on exam for an objective, then 6 questions are in bank.

Exam development and review guidelines: Key should contain the following:

o TR-AA-230-1003, SAT Development Learning Objective Number Test Item o Question or Statement o All possible answers o Correct Answer Indicated o Point Value o References (if applicable)

NOTE: NRC exams may require additional information. Refer to fleet and site specific procedures.

EXAMINATION REVIEW AND APPROVAL:

Developed by: Date:

Instructional Review of Written Exam (Qualified Instructor): Date:

Technical Review (SME): Date:

Approved by Training Supervisor: Date:

Approved by Training Program Owner (or line designee): Date:

Indicate in the following table if any changes are made to the exam after approval:

AR/TWR# PREPARER DATE

DESCRIPTION OF CHANGE REASON FOR CHANGE (if applicable) SUPERVISOR DATE Filename: 50007_PDA 15-1 RO NRC Written_xm TR-AA-230-1003-F13 Revision 1

WRITTEN / ORAL EXAMINATION COVER SHEET Page 1 Trainee Name:

Employee Number: Site: DAEC Examination Number/Title: PDA 15-1 RO NRC Written Exam Training Program: Licensed Operator Initial Training Course/Lesson Plan Number(s): 50007 Reactor Operator Total Points Possible: 75 PASS CRITERIA: 80%. Grade: /75= %

Graded by: Date:

Co-graded by (if necessary): Date:

EXAMINATION RULES

1. References may not be used during this examination, unless otherwise stated.

2. Read each question carefully before answering. If you have any questions or need clarification during the examination, contact the examination proctor.

3. Conversation with other trainees during the examination is prohibited.

4. Partial credit will not be considered, unless otherwise stated. Show all work and state all assumptions when partial credit may be given.

5. Rest room trips are limited and only one examinee at a time may leave.

6. For exams with time limits, you have 360 minutes to complete the examination.

7. Feedback on this exam may be documented on TR-AA-230-1004-F03, Examination Feedback Form. Contact Instructor to obtain a copy of the form.

EXAMINATION INTEGRITY STATEMENT Cheating or compromising the exam will result in disciplinary actions up to and including termination.

I acknowledge that I am aware of the Examination Rules stated above. Further, I have not given, received, or observed any aid or information regarding this examination prior to or during its administration that could compromise this examination.

Examinees Signature: Date:

REVIEW ACKNOWLEDGEMENT I acknowledge that the correct answers to the exam questions were indicated to me following the completion of the exam. I have had the opportunity to review the examination questions with the instructor to ensure my understanding.

Examinees Signature: Date:

Filename: 50007_PDA 15-1 RO NRC Written_xm TR-AA-230-1003-F12 Revision 1

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295001 A2.05 Importance Rating Partial or Complete Loss of Forced Core Flow Circulation: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Jet pump operability. (41.10)

Proposed Question: RO Question # 1 Which of the following conditions would be indicative of a failed inlet elbow (rams head) hold down bolt that AFFECTS reactor recirculation jet pumps?

The rams head has become disconnected Core Plate Differential Pressure will ___(1)___ and the affected recirculation pump discharge flow will ___(2)___.

A. (1) Decrease (2) Increase B. (1) Increase (2) Increase C. (1) Increase (2) Decrease D. (1) Decrease (2) Decrease 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 2

Proposed Answer: A Explanation (Optional):

A. Correct.

B. Incorrect: a decrease in core plate differential pressure indication would occur as a result of a failed recirculation jet pump.

C. Incorrect: a decrease in core plate differential pressure indication would occur as a result of a failed recirculation jet pump. Also, the flow indication for the recirculation loop associated with the failed jet pump would increase.

D. Incorrect: the flow indication for the recirculation loop associated with the failed jet pump would increase.

Technical Reference(s): SD-264, Revision 13 SD-262, Revision 7 Proposed References to be provided to applicants during examination: N Learning Objective: 12.00.00.08 (As available)

Question Source: Bank # DAEC Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 3

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295003 2.1.25 Importance Rating Ability to interpret reference materials, such as graphs, curves, tables, etc. Partial or Complete Loss of A.C. Power.

Proposed Question: RO Question # 2 Given the following:

The plant is operating at 100% power

The M CB8490 Breaker is open and unavailable due to ITC Maintenance

A lightning strike in the switchyard causes the J CB5550 and K CB5560 Breakers to TRIP and Lockout What is the expected plant response 5 minutes after the lightning strike?

A. 1A1, 1A2, 1A3 and 1A4 are de-energized.

B. 1A1, 1A2, 1A3 and 1A4 are energized from 1X3, Startup Transformer.

C. 1A1 and 1A2 are de-energized and 1A3 and 1A4 are energized from the Standby Diesel Generators.

D. 1A1 and 1A2 are on 1X2, Aux Transformer and 1A3 and 1A4 are on 1X4, the Standby Transformer.

Proposed Answer: C Explanation (Optional):

A. Incorrect: the Standby Diesel Generators will repower 1A3 and 14A after +/-10 seconds.

It must be understood that a lockout of the K breaker will not lockout the essential buses and that there is not an electrical path from the J breaker.

B. Incorrect: the Standby Transformer will be de-energized due to the loss of power through the J and K breaker.

C. Correct: this is the expected response due to the loss of power to the essential buses causing a loss of power to RPS and a Group 1-5 isolation closing the MSIVs. This will cause a loss of steam to the turbine preventing the Aux transformer from having power.

The transfer to the Startup Transformer will not occur due to it being de-energized.

D. Incorrect: the Aux Transformer will not be energized due to the turbine tripping on reverse power following the scram caused by the loss of RPS. There is a flow path through the Main Transformer to the Aux Transformer that locks out on the turbine trip that may be used in a backfeed during outage conditions, but is prevented with a backup lockout on the turbine trip.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 4

Technical Reference(s): AOP-301.1, Revision 55 SD-358, Revision 9 SD-304, Revision 19 Switchyard and Bus Proposed References to be provided to applicants during examination:

Drawing Learning Objective: 14.00.00.04, 15.00.00.04 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 5

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295004 K1.03 Importance Rating Knowledge of the operational implications of the following concepts as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Electrical bus divisional separation. CFR: 41.8 to 41.10 Proposed Question: RO Question # 3 Given the following:

The Plant was operating at 100% power THEN:

1C08A (A-9),125V DC System 1 Trouble was received

125V DC System 1 voltage is reported to be zero (0) volts What indication will the operator see in the Control Room for A Condensate Pump 1P-8A?

A. Red light illuminated, pump is running B. No light illuminated, pump is stopped C. Annunciator 1C06A(A-12), "A" Condensate Pump 1P-8A Trip or Motor Overload, in alarm, pump is stopped D. No light illuminated, pump is running Proposed Answer: D Explanation (Optional):

A. Incorrect: the pump has lost control power and will have no indication.

B. Incorrect: the pump remains running and due to the loss of control power will have no indication.

C. Incorrect: ARP 1C06A (A-12) illuminates if an automatic pump trip occurs. Although the pump will display no running indication it has not tripped however, and remains running.

D. Correct.

Technical Reference(s): SD-304, Revision 19 SD-375, Revision 8 SD-639, Revision 9 ARP 1C08A Revision 86 AOP 302.1, Revision 54 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 6

Proposed References to be provided to applicants during examination: N Learning Objective: 13.00.00.04 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 7

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295005 K2.08 Importance Rating Knowledge of the interrelations between MAIN TURBINE GENERATOR TRIP and the following: A.C. electrical distribution. CFR: 41.7 Proposed Question: RO Question # 4 Given the following:

The plant was operating at 25% power during a startup

Following the transfer of nonessential buses 1A1 and 1A2 from the Startup to the Auxiliary Transformer, 1C08B(C-5), Auxiliary XFMR 1X2 Trouble, was received

Immediately thereafter, an electrical fault in the Auxiliary Transformer resulted in the trip of the Generator Backup Lockout Relay FIVE (5) minutes after this fault the Main Turbine has ___(1)___. Non-essential buses 1A1 and 1A2 have experienced a(n) ___(2)___.

A. (1) tripped (2) closed circuit auto transfer B. (1) tripped (2) open circuit auto transfer C. (1) NOT tripped (2) closed circuit auto transfer D. (1) NOT tripped (2) open circuit auto transfer Proposed Answer: B Explanation (Optional):

A. Incorrect; for a closed circuit (make-before-break) transfer for non-essential buses 1A1 and 1A2 to occur, Main Generator Lockout Relays 286/P and 286/B must both be reset.

If either lockout relay is energized (as in the case of the fault in this question), an open circuit (break-before-make) bus transfer will occur.

B. Correct.

C. Incorrect; the 286BU relay duplicates the functions of the 286P relay and will initiate a turbine trip. Also, for a closed circuit (make-before-break) transfer for non-essential buses 1A1 and 1A2 to occur, Main Generator Lockout Relays 286/P and 286/B must both be reset. If either lockout relay is energized (as in the case of the fault in this 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 8

question), an open circuit (break-before-make) bus transfer will occur.

D. Incorrect; the 286BU relay duplicates the function of 286P relay and will initiate a Main Turbine trip.

Technical Reference(s): SD-304, Revision 19 Proposed References to be provided to applicants during examination: N Learning Objective: 14.00.00.03 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 9

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295006 AK 3.06 Importance Rating Knowledge of the reasons for the following responses as they apply to SCRAM : Recirculation pump speed reduction. CFR: 41.5 Proposed Question: RO Question # 5 Given the following:

The plant was operating at 100% reactor power

An automatic scram occurred due to a failure of the pressure regulating system

The crew has stabilized the plant in accordance with EOP 1, RPV Control

RPV water level is being automatically controlled with a Startup FRV and one feed pump in operation

PR-4563, Reactor Pressure Recorder, recorded a maximum pressure of 1120 psig

RPV pressure is being controlled automatically with Low-Low Set When the plant stabilizes following the transient, the Reactor Recirculation pumps will .

A. be operating at approximately 45% speed to ensure that reactor power output does not exceed the capacity of the operating feed pump B. have tripped due to actuation of the ATWS - RPT logic on high RPV pressure C. be operating at minimum speed to prevent cavitation of the jet pumps resulting from a reduction in NPSH D. have tripped to ensure that MCPR limits are not exceeded due to the pressure transient caused by the closure of the main turbine control valves Proposed Answer: C Explanation (Optional):

A. Incorrect - Plausible since RR pumps do runback to 45% on the trip of one feed pump but the 20% limiter will be in operation due to low feed flow.

B. Incorrect - ATWS RPT does not actuate until RPV pressure rises above 1140 psig. LLS Reliefs cycle between 910 and 1035 psig after be armed at 1110 psig (PSV-4401 relief setpoint) with a confirmatory Reach High Pressure Trip of 1055 psig.

C. Correct - RR pumps runback to 20% when feed flow is less than 20% to prevent cavitation of the pumps due to the reduction in NPSH D. Incorrect - EOC-RPT is actuated by fast closure (low ETS oil pressure of 800 psig) of the TCVs or closure of the TSVs. Valves operated normally, albeit in response to the control system failure.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 10

Technical Reference(s): SD-264, Rev 13 Proposed References to be provided to applicants during examination: N Learning Objective: 12.00.00.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 11

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295016 AA1.04 Importance Rating Ability to operate and or monitor the following responses as they apply to Control Room Abandonment: A.C. Electrical Distribution.

Proposed Question: RO Question # 6 Given the following:

The plant is operating at 100% reactor power

A fire has occurred INSIDE 1C08

AOP 913, Fire, has been entered

AOP 915, Shutdown Outside of the Control Room, has been entered

Control has been shifted to 1C388

Subsequently, a loss of offsite power has occurred Which one of the following is correct regarding the B SBDG and re-energizing bus 1A4?

In accordance with AOP 915, the B SBDG will ___(1)___ and the B SBDG output breaker will need to be closed in from ___(2)___.

A. (1) automatically start (2) the 1A4 Switchgear Room B. (1) automatically start (2) 1C-94 in the B SBDG Room C. (1) need to be started from 1C-94 in the B SBDG Room (2) 1C-94 in the B SBDG Room D. (1) need to be started from 1C-94 in the B SBDG Room (2) 1C388, Remote Shutdown Panel Proposed Answer: D Explanation (Optional):

A. Incorrect - AOP 915 contains a note on page 52 which states that the SBDG 1G-21 will not automatically start and power bus 1A4 on a loss of power when control has been shifted to 1C388. AOP 915, Section 7, Step 2e and f, states that the 1G-21 must be started from 1C-94, in the SBDG Room, and the SBDG output breaker is closed at 1C388.

B. Incorrect - AOP 915 contains a note on page 52 which states that the SBDG 1G-21 will not automatically start and power bus 1A4 on a loss of power when control has been 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 12

shifted to 1C388.

C. Incorrect - AOP 915, Section 7, Step 2e and f, states that the 1G-21 must be started from 1C-94, in the SBDG Room, and the SBDG output breaker is closed at 1C388.

D. Correct - AOP 915 contains a note on page 52 which states that the SBDG 1G-21 will not automatically start and power bus 1A4 on a loss of power when control has been shifted to 1C388. AOP 915, Section 7, Step 2e and f on page 55, states that the 1G-21 must be started from 1C-94, in the SBDG Room, and the SBDG output breaker is closed at 1C388.

Technical Reference(s): AOP 915, Rev. 53 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 13

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295018 A2.02 Importance Rating Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF COMPONENT COOLING WATER: Cooling water temperature. CFR: 41.10 Proposed Question: RO Question # 7 Given the following:

The plant is in Mode 1 at 95% power

Reactor Building Closed Cooling Water Heat Exchanger 1E-35A was isolated due to a valve lineup error Which of the following describes a potential consequence with respect to the reactor recirculation system if this condition is left uncorrected?

A. Reactor recirculation pump stator winding insulation degradation B. Recirculation Pump motor bearing damage C. Recirculation Pump M-G set generator winding degradation D. Recirculation Pump M-G set motor bearing damage Proposed Answer: B Explanation (Optional):

A. Incorrect; Recirculation Pump stator windings are air cooled. Rising RBCCW system temperature will not cause an adverse effect.

B. Correct; Recirculation Pump lubricating oil is cooled by RBBCW. As RBBCW system temperature rises, so will the oil temperature for the recirculation pumps. High oil temperature can result in motor bearing damage due to a reduction in lubrication.

C. Incorrect; Recirculation Pump M-G set generators are air cooled. Rising RBCCW system temperature will not cause an adverse effect.

D. Incorrect; Recirculation Pump M-G set oil coolers are cooled by GSW, not RBBCW.

Rising RBCCW system temperature will not cause an adverse effect.

Technical Reference(s): SD-264, Revision 13 SD-414, Revision 9 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 14

Proposed References to be provided to applicants during examination: N Learning Objective: 29.00.00.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 15

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295019 2.2.44 Importance Rating (Partial or Total Loss of Inst. Air): Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. CFR: 41.5 Proposed Question: RO Question # 8 Given the following:

The plant is operating at 85% power

Five minutes ago an air leak occurred in the Instrument Air common supply piping located downstream of the Instrument Air Dryers

AOP-518, Failure of Instrument and Service Air, has been entered

CV-3034, Balance of Plant Instrument Air Header Isolation Valve, has been verified closed

Instrument Air Pressure is currently 78 psig Which of the following describes how Main Feedwater Regulating Valves would operate based on the existing plant conditions?

A. operate normally in response to control signals B. not be able to move due to locking up C. will be drifting to an open position D. will be drifting to a closed position Proposed Answer: A Explanation (Optional):

A. Correct: Balance of Plant Instrument Air Header Isolation Valve CV 3034 would have closed at 80 psig, isolating the normal Instrument Air supply to the FRVs. CV-1579, A Feed Regulating Valve, and CV-1621, B Feed Regulating Valve, have backup air accumulators that will provide approximately 30 minutes of continuous operation as discussed in AOP-518. Operator action to throttle A and B Feedline Block valves MO 1592 and MO 1636 is not required yet.

B. Incorrect: The lockout relay actuates at an air supply pressure of <75 psig as sensed by the supply pressure switch. Pressure is still above this setpoint.

C. Incorrect: AOP-518 states that during a prolonged loss of air casualty the Feed Reg valves may drift open. Additionally, backup accumulators provide for 30 minutes of continuous operation. This question assumes only five minutes have elapsed.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 16

D. Incorrect: AOP-518 states that during a prolonged loss of air casualty the FRVs may drift open. The failure direction of the valves would be open, not closed.

Technical Reference(s): SD-518, Revision 9 SD-644, Revision 14 AOP-518, Revision 34 Proposed References to be provided to applicants during examination: N Learning Objective: 45.00.00.03 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 17

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295021 K1.01 Importance Rating Knowledge of the operational implications of the following concepts as they apply to LOSS OF SHUTDOWN COOLING: Decay heat. CFR: 41.8 to 41.10 Proposed Question: RO Question # 9 Given the following:

The reactor has been shutdown for 10 days

RPV level is 200" and steady

Shutdown cooling has just been lost

Reactor coolant temperature is currently 100°F and rising

Reactor recirculation pumps are not available With the given conditions:

(1) What is the approximate TIME it will take to reach boiling in the reactor?

AND (2) What alternate decay heat removal method will remove the MOST decay heat?

A. (1) 4.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months (2) Feed and Bleed to Radwaste or Condenser B. (1) 4.7 hours8.101852e-5 days 0.00194 hours 1.157407e-5 weeks 2.6635e-6 months (2) Reactor Water Cleanup Heat Exchanger C. (1) 24.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months (2) Feed and Bleed to Radwaste or Condenser D. (1) 24.9 hours1.041667e-4 days 0.0025 hours 1.488095e-5 weeks 3.4245e-6 months (2) Reactor Water Cleanup Heat Exchanger Proposed Answer: A Explanation (Optional):

A. Correct: if the correct graph is used (APPENDIX 2 - HEATUP RATE CURVE RPV LEVEL AT 200"), an approximate heatup rate of 24°F/hr will be obtained. (212°F -

100°F) / (24°F/hr) = 4.7°F/hr. Feed and Bleed to Radwaste or Condenser meets the Tech Spec requirements for alternate decay heat removal (AOP 149 page 7).

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 18

B. Incorrect: if the correct graph is used (APPENDIX 2 - HEATUP RATE CURVE RPV LEVEL AT 200"), an approximate heatup rate of 24°F/hr will be obtained. (212°F -

100°F) / (24°F/hr) = 4.7°F/hr. The reactor water cleanup heat exchanger system has a low net heat removal rate and should only be used if RWCU is not needed as a drain path for a feed and bleed operation or in circumstances where the bulk coolant temperatures are low enough that feed and bleed operations do not remove significant heat (AOP-149 page 9).

C. Incorrect: this answer would be reached if the incorrect graph was used (APPENDIX 1 -

HEATUP RATE CURVE - RPV FLOODED). This would yield and approximate heatup rate of 4.5°F/hr. (212°F - 100°F) / (4.5°F/hr) = 24.9°F/hr. Feed and Bleed to Radwaste or Condenser meets the Tech Spec requirements for alternate decay heat removal (AOP 149 page 7).

D. Incorrect: this answer would be reached if the incorrect graph was used (APPENDIX 1 -

HEATUP RATE CURVE - RPV FLOODED). This would yield and approximate heatup rate of 4.5°F/hr. (212°F - 100°F) / (4.5°F/hr) = 24.9°F/hr. The reactor water cleanup heat exchanger system has a low net heat removal rate and should only be used if RWCU is not needed as a drain path for a feed and bleed operation or in circumstances where the bulk coolant temperatures are low enough that feed and bleed operations do not remove significant heat (AOP-149 page 9)

Technical Reference(s): AOP-149, Revision 41 Proposed References to be provided to applicants during examination: Y AOP-149, Appendix 1 (HEATUP RATE CURVE - RPV FLOODED) and Appendix 2 (HEATUP RATE CURVE - RPV LEVEL AT 200"), without CAUTIONS or NOTES (graphs ONLY).

Learning Objective: 94.01.02.03 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 19

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295023 Importance Rating Knowledge of the interrelations between REFUELING ACCIDENTS and the following: Fuel handling equipment. CFR: 41.7 Proposed Question: RO Question # 10 Consider the following plant configuration:

Mode Switch is in REFUEL

All control rods are fully inserted

One control rod is selected

There are no Rod Blocks in effect

The only hoist in use is the Main Grapple Hoist

A fuel assembly is grappled and raised to Full-Up in the Fuel Pool

The Refuel Platform is driven over the core

The fuel assembly is lowered into its assigned location

The 1C05 operator attempts to withdraw the selected control rod At which point in the following Core Alteration scenario would activation of the annunciator 1C05B(A-6), ROD Out Block, FIRST occur?

A. Refuel Platform is driven over the core B. Main Grapple Hoist starts to lower the fuel assembly C. Main Grapple Hoist reaches Full-Up in the Fuel Pool D. 1C05 operator attempts to withdraw the selected control rod Proposed Answer: A Explanation (Optional):

A. Correct - The following Rod blocks occur with Refuel Platform over the Core and mode switch in Refuel:

Mode Switch in Refuel AND Refuel Platform over the Core AND A. Frame Mounted Hoist loaded > 400 lbs Or B. Trolley mounted Hoist loaded > 400 lbs Or C. Fuel Grapple loaded > 400 lbs Or D. Fuel Grapple not full up B. Rod Block occurs earlier, when Platform is moved over the core C. Does not cause a rod block because the platform has not been moved over the core 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 20

D. Rod Block occurs earlier, when Platform is moved over the core Technical Reference(s): SD-281, Rev 7 1C05B(A-6), Rev 98 Proposed References to be provided to applicants during examination: N Learning Objective: 98.03.01.05 (As available)

Question Source: Bank # 19677 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 21

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295024 K3.08 Importance Rating Knowledge of the reasons for the following responses as they apply to HIGH DRYWELL PRESSURE: Containment Spray CFR: 41.5 Proposed Question: RO Question # 11 Which of the following describes the reason that initiation of Drywell Spray is permitted only within the limits of the Drywell Spray Initiation Curve (DWSIL)?

A. To ensure that Suppression Chamber Pressure can be restored below the Torus Spray Initiation Pressure B. To ensure that cycling of the reactor building to torus vacuum breakers is minimized and to prevent challenges of the primary containment pressure suppression capability C. To prevent an evaporative cooling pressure drop large enough to challenge containment integrity or draw in air through the torus to drywell vacuum breakers D. To prevent an evaporative cooling pressure drop large enough to challenge containment integrity or draw in air through the reactor building to torus vacuum breakers Proposed Answer: D Explanation (Optional):

A. Incorrect: Restoring drywell pressure below the Torus Spray Initiation Pressure may occur, but is not the basis for DWSIL.

B. Incorrect. Cycling of the breakers is not a concern in the shaded area of the curve.

C. Incorrect. The concern is with de-inerting the primary containment atmosphere through the reactor building to torus vacuum breakers. The torus to drywell vacuum breakers are within the containment atmosphere.

D. Correct. Unrestricted spray operation could result in a negative pressure large enough to de-inert the primary containment or challenge the primary containment negative pressure capability.

Bases-EOP CURVES AND Technical Reference(s):

LIMITS, Rev. 13 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 22

Proposed References to be provided to applicants during examination: N Learning Objective: 95.00.00.03 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 23

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating KA295025 High Reactor Pressure Ability to operate and/or monitor the following as they apply to HIGH REACTOR PRESSURE:

Reactor/turbine pressure regulating system. CFR: 41.7 Proposed Question: RO Question # 12 Give the following:

Reactor power is lowered to 65% to isolate A Inboard MSIV. Plant conditions are:

No adjustments have been made to EHC Pressure Set

Reactor pressure is now 990 psig

Main generator MW output is 415 MW When A Inboard MSIV is isolated in accordance with OI 683, Main Steam System:

(1) What is the expected reactor pressure response?

AND (2) What is the expected reactor power change?

A. (1) Reactor pressure will remain at 990 psig (2) Reactor power will lower B. (1) Reactor pressure will lower below 990 psig (2) Reactor power will lower C. (1) Reactor pressure will remain at 990 psig (2) Reactor power will rise D. (1) Reactor pressure will raise greater than 990 psig (2) Reactor power will rise Proposed Answer: D Explanation (Optional):

A. Incorrect - Reactor pressure will rise due to increased head loss with A MSIV being closed.

B. Incorrect - Reactor pressure will rise due to increased head loss with A MSIV being closed.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 24

C. Incorrect - Reactor pressure will rise due to increased head loss with A MSIV being closed.

D. Correct - Reactor pressure will raise to an approximate value greater than the indicated reactor pressure provided in the STEM due to increased head loss with A MSIV being closed. Reactor power will rise due to reactor pressure rising collapsing voids and adding positive reactivity.

Technical Reference(s): IPOI 3, Rev. 143 683, Rev. 53 SD-683, Rev. 8 Proposed References to be provided to applicants during examination: N Learning Objective: 48.03.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 25

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295026 EA 2.03 Importance Rating 3.9 Ability to determine and/or interpret the following as they apply to SUPPRESSION POOL HIGH WATER TEMPERATURE: Reactor Pressure Proposed Question: RO Question # 13 Given the following plant conditions:

The crew is performing actions of EOP ATWS, RPV Control

Boron injection using SBLC has been successfully initiated

RPV water level is being controlled between -25 and +15 with Feedwater

Before Defeat 15 could be implemented, MSIVs closed on RPV Lo-Lo-Lo Level

RPV pressure is controlled 800 to 1000 psig with manual operation of an SRV

Reactor Power remains at approximately 10%

Torus water temperature is 130°F and rising 1°F per minute with both loops of RHR operating in the Torus Cooling mode

Torus water level is 10.3 ft. and steady

Defeat 15 is currently installed (NOTE: this is additional information provided by NRC during exam in response to a question by an applicant).

With the provided conditions, which ONE of the following RPV pressure control strategies is correct?

A. Immediately establish an RPV pressure band of 400 to 600 psig B. WAIT until the Cold Shutdown Boron Weight has been injected, THEN establish an RPV pressure band of 400 to 600 psig C. Re-open the MSIVs and use the main turbine bypass valves to rapidly depressurize the RPV irrespective of cooldown limits D. Re-open the MSIVs and use the main turbine bypass valves and/or main steam line drains to control pressure 800 to 1000 psig Proposed Answer: D Explanation (Optional):

A. Incorrect - Plausible since lowering the pressure control band will initially increase the margin HCL. While lowering the pressure control band will delay exceeding HCL, it will not stop heat addition to the Torus. With the current control band and the current heatup rate of the Torus, HCL would not be exceed for 20-30 minutes. This must be balanced the potential for power increase from cooling down.

B. Incorrect - Plausible since lowering the pressure control band will initially increase the margin HCL and normally cooldown is not permitted until boron addition is complete. It 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 26

is unnecessary to wait for boron addition to be completed before lowering pressure to prevent exceeding HCL (refer to override in Pressure Control Leg prior to step P-4.

Additionally, While lowering the pressure control band will delay exceeding HCL, it will not stop heat addition to the Torus. With the current control band and the current heatup rate of the Torus, HCL would not be exceed for 20-30 minutes. This must be balanced the potential for power increase from cooling down.

C. Incorrect - It is not permissible to anticipate blowdown while implementing the ATWS -

RPV Control EOP. Plausible since anticipating blowdown is preferable to ED in non-ATWS scenarios.

D. Correct -- Re-opening the MSIVs and transferring heat to the condenser instead of the Torus is much more desirable. Refer to override in Pressure Control Leg prior to step P-4.

Technical Reference(s): Bases ATWS Rev 17 Proposed References to be provided to applicants during examination: EOP Graph 4, HCL Learning Objective: 95.00.00.03 (As available)

Question Source: Bank #

Modified Bank # 39477 (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 27

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295028 EK1.01 Importance Rating Knowledge of the operational implications of the following concepts as they apply to HIGH DRYWELL TEMPERATURE : Reactor water level measurement. (CFR: 41.8 to 41.10)

Proposed Question: RO Question # 14 Which of the following explains the concern with the Minimum Indicating Level (MIL) for the RPV water level instruments?

With elevated drywell temperatures.

A. ALL RPV level instruments may indicate a level even when their REFERENCE leg tap is uncovered B. ALL RPV level instruments may indicate a level even when their VARIABLE leg tap is uncovered C. WR Yarway and Floodup RPV level instruments may indicate a level when their REFERENCE leg tap is uncovered D. WR Yarway and Floodup RPV level instruments may indicate a level when their VARIABLE leg tap is uncovered Proposed Answer: D Explanation (Optional):

A. Incorrect: DAEC RPV water level instruments sense level by measuring the differential pressure (P) between a reference leg water column and a variable leg water column.

The reference leg is kept full of water by a condensing pot replenished with steam from the RPV. The variable leg height depends on RPV water level. When the actual RPV water level decreases, the variable leg height also decreases, causing the sensed P to increase. The higher P results in a lower indicated level. By design, the reference leg remains uncovered. If the reference leg becomes covered, no P is sensed, indicated RPV water level would display maximum indicated water level.

B. Incorrect: Most of the Narrow Range GEMAC and Fuel Zone instrument runs outside the drywell. With the actual RPV water level at the elevation of the variable leg tap, these instruments will read on-scale only at relatively high reactor building temperatures. MILs are therefore unnecessary for the Narrow Range GEMAC and Fuel Zone instruments; as long as the indicated level is on-scale, the actual level must be above the variable leg tap and the instruments can be used to evaluate the level trend.

C. Incorrect: DAEC RPV water level instruments sense level by measuring the P between 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 28

a reference leg water column and a variable leg water column. The reference leg is kept full of water by a condensing pot replenished with steam from the RPV. The variable leg height depends on RPV water level. When the actual RPV water level decreases, the variable leg height also decreases, causing the sensed P to increase. The higher P results in a lower indicated level. By design, the reference leg remains uncovered. If the reference leg becomes covered, no P is sensed, indicated RPV water level would display maximum indicated water level.

D. Correct: With actual RPV water level at the elevation of the variable leg tap, the instrument should read downscale low. As drywell temperature rises, however, the resulting change in the density of water in the instrument runs decreases the P between the variable and reference legs. At a drywell temperature of 158 F the change in P is sufficient to begin to drive the indicated level on-scale. DCP-1410 re-routed most of the Narrow Range GEMAC and Fuel Zone instrument runs outside the drywell, therefore only the WR Yarway and Floodup RPV level instruments have MILs outlined in EOP Curves and Limits, Caution #1.

Technical Reference(s): EOP Curves and Limits, Rev. 13 Proposed References to be provided to applicants during examination: N Learning Objective: 88.00.00.06 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 29

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295030 EA 1.05 Importance Rating Ability to operate and/or monitor the following as they apply to LOW SUPPRESSION POOL WATER LEVEL: HPCI Proposed Question: RO Question # 15 Current plant conditions are as follows:

EOP-1, RPV Control is being executed

High Pressure Coolant Injection (HPCI) is operating for RPV level control per EOP 1, RPV Control

Torus Water Level is reported to be 10.1 feet and LOWERING Which of the following identifies the Torus water level at which HPCI must be secured AND the reason it must be secured?

HPCI must be secured when Torus water level reaches .

A. 7.1 feet, to prevent violating Vortex Limits B. 7.1 feet, to prevent direct pressurization of the Torus by the HPCI exhaust C. 5.8 feet, to prevent violating Vortex Limits D. 5.8 feet, to prevent direct pressurization of the Torus by the HPCI exhaust Proposed Answer: D Explanation (Optional):

A. Incorrect - This is a vortex level limit UNLESS directed to use HPCI in EOPs B. Incorrect - This is above the level that will result in direct pressurization of the torus by the HPCI exhaust.

C. Incorrect - EOP 1 overrides vortex concerns D. Correct - Per EOP 2 bases step T/L A torus level of 5.8 feet corresponds to the HPCI turbine exhaust elevation. Direction here attempts to maintain the availability of HPCI should it be needed as an injection source or alternate method of depressurizing the RPV. Operation of the HPCI system with its exhaust device not submerged will directly pressurize the torus.

Technical Reference(s): Bases-EOP 1, Rev. 16 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 30

Proposed References to be provided to applicants during examination: N Learning Objective: 95.00.00.03 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 31

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295031 K3.01 Importance Rating Knowledge of the reasons for the following responses as they apply to REACTOR LOW WATER LEVEL: Automatic depressurization system actuation. CFR: 41.5 Proposed Question: RO Question # 16 Given the following:

The plant was operating at 99% reactor power when rising drywell pressure and lowering Reactor Pressure Vessel (RPV) level resulted in the crew initiating a manual scram

High Pressure Coolant Injection (HPCI) flow could not be established

All Residual Heat Removal (RHR) and Core Spray Pumps are running

1C03A(C-6) ADS/LLS 125 VDC Control Power Failure, IS in alarm due to a loss of normal 125 VDC power to Automatic Depressurization System (ADS) logic B

1C03A (A-7), ADS LO Water Level Confirmed, is NOT in alarm

RPV water level has just reached 64 inches and continues to lower Which of the following describes the expected response of the Automatic Depressurization System 120 seconds later?

A. ADS logic train A and B will open FOUR ADS valves B. ADS logic train A will open FOUR ADS valves C. ADS logic train A will open TWO ADS valves D. ADS logic train A and B will NOT open ANY ADS valves Proposed Answer: D Explanation (Optional):

A. Incorrect: Loss of normal 125 VDC power to ADS logic B results in a transfer to the alternate 125 VDC power source. However, ADS will not initiate without the confirmatory 170 signal present (1C03A (A-7), ADS LO WATER LEVEL CONFIRMED is extinguished). Additionally, the ADS 2 minute timers will not actuate (1C03A (A-5),

ADS A/B 2 MIN Tier(s) Initiated, will not alarm).

B. Incorrect: Either ADS logic train will open all four ADS valves. However, ADS will not initiate without the confirmatory 170 signal present.

C. Incorrect: Either ADS logic train will open all four ADS valves. Loss of DC power to one logic train would not cause a half-actuation of the system.

D. Correct: Loss of normal 125 VDC power to ADS logic B results in a transfer to the alternate 125 VDC power source, thus ADS is capable of initiating. The required low-50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 32

low-low RPV water level has been reached and the required pumps (RHR and/or Core Spray) are running. However, ADS will not initiate without the confirmatory 170 signal present (this information is provided by the 1C03A (A-7), ADS LO WATER LEVEL CONFIRMED, is NOT lit in the stem). Additionally, the ADS 2 minute timers will not actuate (1C03A (A-5), ADS A/B 2 MIN Tier(s) Initiated, will not alarm).

Technical Reference(s): SD-183.1, Revision 7 ARP 1C03A (A-7)

ARP 1C03A (C-6)

Proposed References to be provided to applicants during examination: N Learning Objective: 8.02.01.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 33

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295037 A1.04 Importance Rating Ability to operate and/or monitor the following as they apply to SCRAM CONDITION PRESENT AND REACTOR POWER ABOVE APRM DOWNSCALE OR UNKNOWN : SBLC. CFR: 41.7 Proposed Question: RO Question # 17 Given the following:

The plant was operating at 40% when a turbine trip occurred

The scram signal failed to result in control rod insertion

EOP 1, RPV Control, has been exited to ATWS, RPV Control

Current reactor power is 12%

Current torus water temperature is 111°F and rising Which of the following describes the BASIS for Standby Liquid Control boron injection?

A. required to avoid exceeding the torus water Heat Capacity Limit B. required to offset voiding for RPV depressurization C. required to offset water density changes for RPV cool down D. NOT yet required Proposed Answer: A Explanation (Optional):

A. Correct: SBLC needs to be initiated. It is desired to shut down the reactor prior to depressurization, and depressurization must occur before the Heat Capacity Limit is reached. The action to initiate SBLC must occur prior to exceeding the limit of EOP Graph 6, Boron Injection Initiation Temperature (BIIT). (SD-153 pages 20 - 24).

B. Incorrect: while offsetting the reactivity effects of changes in voiding is a design basis for the SBLC system (SD-153 pages 4-5), and a depressurization will subsequently occur, the required initiation of SBLC with regard to EOP Graph 6 is driven by concern for exceeding the Heat Capacity Limit.

C. Incorrect: while offsetting the reactivity effects of changes in water density is a design basis for the SBLC system (SD-153 pages 4-5), and a cooldown will subsequently occur, the required initiation of SBLC with regard to EOP Graph 6 is driven by concern for exceeding the Heat Capacity Limit.

D. Incorrect: SBLC boron injection is required prior to torus water temperature exceeding the limit curve of EOP Graph 6 to prevent exceeding the Heat Capacity Limit. (SD-153 pg. 20 and 24). It must be understood that this action is required prior to exceeding the curve, and not afterwards.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 34

Technical Reference(s): ATWS, Revision 21 EOP 1, Revision 18 EOP Graph 6, Rev. 7 SD-153, Revision 8 Proposed References to be provided to applicants during examination: EOP Graph 6, BIIT Learning Objective: 95.00.00.03 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 35

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295038 EA 2.01 Importance Rating Ability to determine and/or interpret the following as they apply to HIGH OFF-SITE RELEASE RATE: Off-site CFR: 41.10 Proposed Question: RO Question # 18 A transient is occurring that requires the Primary Containment to be vented to maintain Torus Pressure below the Primary Containment Pressure Limit (PCPL).

Torus Pressure is 40 psig and rising slowly

Torus Water Level is 14 feet

Emergency Depressurization has been completed Which one of the following Containment Venting paths should result in the lowest off-site release?

A. Venting the Drywell via the Hard Pipe Vent B. Venting the Torus Air Space via the Hard Pipe Vent C. Venting the Drywell via the 2 vent line to Standby Gas Treatment System D. Venting the Torus Air Space via the 2 vent line to Standby Gas Treatment System Proposed Answer: D Explanation (Optional):

A. Incorrect - while physically possible this path has no procedure for use. Additionally this path would not take advantage of scrubbing by the Torus Pool and is unfiltered.

B. Incorrect - This path does take advantage of scrubbing by the Torus Pool but is not filtered and has a larger diameter vent path permitting higher flow rates.

C. Incorrect - This path is filtered by the Standby Gas Treatment System, but does not take advantage of scrubbing by the Torus Pool D. Correct - This path takes advantage of scrubbing by the Torus Pool and is filtered by the Standby Gas Treatment System.

SEP 301.2, Drywell Vent Via Technical Reference(s): SEP 301.1, Torus Vent Via SBGT SBGT SEP 301.3, Torus Vent Technical Support Guideline Via Hard Pipe Vent Bases-EOP 2, Rev. Appendix C, Containment Venting 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 36

Proposed References to be provided to applicants during examination: N Learning Objective: 95.00.00.03 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 37

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 600000 2.4.45 Importance Rating (Plant Fire On Site): Ability to prioritize and interpret the significance of each annunciator or alarm. CFR: 41.10 Proposed Question: RO Question # 19 The following annunciators are received on 1C-40:

In accordance with AOP 913, Fire, what action(s) is(are) required?

A. Activate the fire brigade B. Start the Diesel Driven Fire Pump 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 38

C. Direct the NSPEO to investigate, fire brigade activation is NOT required D. Contact offsite fire assistance, fire brigade activation is NOT required Proposed Answer: A Explanation (Optional):

A. Correct - AOP 913, Step 3 on page 3, states that if any of the following conditions exist, then activate the fire brigade: A fire alarm in conjunction with the following:

sprinkler/deluge initiation or any fire pump(s) auto starting.

B. Incorrect - AOP 913, Step 7 on page 3, states if fire water is required for firefighting, then verify 1P-48 Electric Fire Pump or 1P-49 Diesel Fire Pump running. Start pumps as required from 1C40. From the STEM of the question, the Electric Fire Pump has already started.

C. Incorrect - AOP 913, Step 3 on page 3, states that if any of the following conditions exist, then activate the fire brigade: A fire alarm in conjunction with the following:

sprinkler/deluge initiation or any fire pump(s) auto starting. From the STEM of the question, the fire brigade is required to be activated.

D. Incorrect - AOP 913, Step 3 on page 3, states that if any of the following conditions exist, then activate the fire brigade: A fire alarm in conjunction with the following:

sprinkler/deluge initiation or any fire pump(s) auto starting. From the STEM of the question, the fire brigade is required to be activated.

Technical Reference(s): AOP 913, Rev. 77 Proposed References to be provided to applicants during examination: N Learning Objective: 94.25.01.03 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 39

55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 40

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 700000 K1.01 Importance Rating Knowledge of the operational implications of the following concepts as they apply to GENERATOR VOLTAGE AND ELECTRIC GRID DISTURBANCES: Definition of terms: volts, watts, amps, VARs, power factor. CFR: 41.4, 41.5, 41.7, 41.10 Proposed Question: RO Question # 20 Given the following:

The plant is conducting a startup following a refueling outage

The operating crew is preparing to synchronize the main generator to the grid

Currently main generator output frequency is 60.1 Hz and grid frequency is 60.0 Hz

Due to a disturbance in the Main Generator voltage regulator, generator output voltage drops below grid voltage Which of the following describes the expected impact to MW and VARS loading when the main generator is subsequently synchronized to the grid?

A. The main generator will become both a real load and a reactive load for the grid B. The main generator will become a real load, but will supply reactive load, to the grid C. The main generator will supply both real load and reactive load to the grid D. The main generator will supply real load, and become a reactive load, to the grid Proposed Answer: D Explanation (Optional):

A. Incorrect: generator frequency is higher than grid frequency, and thus the generator will assume real (MW) loading.

B. Incorrect: generator frequency is higher than grid frequency, and thus the generator will assume real (MW) loading. Also, since the voltage regulator disturbance has caused generator output voltage to lower below grid voltage, the generator will become a reactive (VARS) load to the grid instead of supplying VARS.

C. Incorrect: since the voltage regulator disturbance has caused generator output voltage to lower below grid voltage, the generator will become a reactive (VARS) load to the grid instead of supplying VARS.

D. Correct: generator frequency is higher than grid frequency, and thus the generator will assume real (MW) loading. Also, since the voltage regulator disturbance has caused generator output voltage to lower below grid voltage, the generator will become a reactive (VARS) load to the grid instead of supplying VARS.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 41

Technical Reference(s): SD-304, Revision 19 SD-698, Revision 5 Proposed References to be provided to applicants during examination: N Learning Objective: 57.00.00.05 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 42

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295008 A1.05 Importance Rating Ability to operate and/or monitor the following as they apply to HIGH REACTOR WATER LEVEL: RCIC. CFR: 41.7 Proposed Question: RO Question # 21 Given the following:

The plant has experienced a station blackout

RCIC is operating and supplying water to the RPV

RPV level has been gradually rising and has just exceeded 211 inches Which of the following will occur DIRECTLY as a result of the conditions described above?

A. MO-2404, RCIC Turbine Steam Supply Isolation, closes B. MO-2405, RCIC Turbine Steam Supply Stop Valve, closes C. MO-2512, RCIC Injection Header Isolation, closes D. MO-2400, RCIC Steam Supply Inboard Isolation, closes Proposed Answer: A Explanation (Optional):

A. Correct: RPV level reaching 211 signals MO-2404 to close (OI-150 page 16)

B. Incorrect: MO-2405 closes in response to RCIC turbine trip signals. RPV high water level does not generate an RPV turbine trip signal. (SD-150 pages 16-18, 23-24).

C. Incorrect: MO-2512 and MO-2404 are interlocked. Closure of MO-2404 causes MO-2512 to close, but only after MO-2404 reaches its shut seat, actuating a limit switch.

MO-2512 is not directly actuated by the 211 high level signal; it is controlled by the action of MO-2404 which operates directly in response to RPV high level at 211 (SD-150 pages 21-22).

D. Incorrect: The receipt of a RCIC isolation signal will result in the closure of both MO-2400 and MO-2401 via the action of separate logic trains. High RPV level is not associated with this RCIC isolation feature and secures RCIC via a separate mechanism (SD-150 pages 15 and 24).

Technical Reference(s): SD-150, Revision 8 OI-150, Revision 77 ARP 1C04C, Rev. 44 AOP 301.1, Revision 55 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 43

Proposed References to be provided to applicants during examination: N Learning Objective: 3.06.01.07 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 44

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295009 A2.02 Importance Rating Ability to determine and/or interpret the following as they apply to LOW REACTOR WATER LEVEL: Steam flow/feed flow mismatch. CFR: 41.10 Proposed Question: RO Question # 22 Given the following:

A plant scram has occurred as a result of Main Steam Isolation Valve closure

The Operator at the Controls (OATC) is controlling level between 150 inches and 211 inches with MANUAL operation of the Startup Feedwater Regulating Valve (FRV)

The OATC is controlling pressure between 800 and 1000 psig by cycling a Safety/Relief Valve (SRV) as necessary

Currently, the following conditions exist:

o RPV level is stable at 170 inches with the Startup FRV CLOSED o RPV pressure is approaching 1000 psig Assuming no additional operator action, what is the expected Reactor Pressure Vessel (RPV) level response when the OATC cycles the SRV open and then closed?

A. rise when the SRV is opened; when the SRV is closed level will lower and stabilize below 170 inches B. rise when the SRV is opened; when the SRV is closed level will stabilize at 170 inches C. initially lower, then rise when the SRV is opened; when the SRV is closed level will stabilize at 170 inches D. lower when the SRV is opened; when the SRV is closed level will stabilize below 170 inches Proposed Answer: A Explanation (Optional):

A. Correct: RPV level rises due to core voiding, then drops to below original level due to inventory removed.

B. Incorrect: loss of inventory through SRV results in final level less than original (170 inches).

C. Incorrect: RPV level rises due to core voiding while SRV is open, then drops to below original level due to inventory removed.

D. Incorrect: RPV level rises due to core voiding while SRV is open.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 45

Technical Reference(s): SD-644, Revision 14 SD-644, Revision 14 Proposed References to be provided to applicants during examination: N Learning Objective: 93.22.01.02 (As available)

(DAEC, source listed as 2004 Question Source: Bank # X River Bend NRC Exam, refer to Main Steam System exam bank)

Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 46

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295014 AK 2.04 Importance Rating Knowledge of the interrelations between INADVERTENT REACTIVITY ADDITION and the following: Void concentration. CFR: 41.7 Proposed Question: RO Question # 23 Given the following:

The plant is operating at 100% reactor power

A spurious main turbine trip has occurred (1) What is the status of the RPT breakers and the MG Set drive motor breakers?

AND (2) What is the reason for this condition?

A. (1) RPT breakers are OPEN and MG Set drive motor breakers are CLOSED (2) lowering core inlet subcooling to limit peak power B. (1) RPT breakers are CLOSED and MG Set drive motor breakers are OPEN (2) lowering core inlet subcooling to limit peak power C. (1) RPT breakers are OPEN and MG Set drive motor breakers are OPEN (2) raise the void fraction in the core to limit peak power D. (1) RPT breakers are OPEN and MG Set drive motor breakers are CLOSED (2) raise the void fraction in the core to limit peak power Proposed Answer: C Explanation (Optional):

A. Incorrect - SD-264, page 15 states that an MG Set drive motor will trip due to the trip of any RPT breaker.

B. Incorrect - SD-264, page 33 states that the RPT feature accomplishes a rapid power reduction due to the rapid reduction of recirculation flow which increases the core void content. The concern of the lower core inlet subcooling is not a design consideration.

C. Correct - SD-264, page 33 states that the RPT feature accomplishes a rapid power reduction due to the rapid reduction of recirculation flow which increases the core void content. SD-264, page 15 states that an MG Set drive motor will trip due to the trip of any RPT breaker.

D. Incorrect - SD-264, page 15 states that an MG Set drive motor will trip due to the trip of 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 47

any RPT breaker. The concern of the lower core inlet subcooling is not a design consideration.

Technical Reference(s): SD-264, Ref. 13 Proposed References to be provided to applicants during examination: N Learning Objective: 93.22.01.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 48

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295020 K1.01 Importance Rating Knowledge of the operational implications of the following concepts as they apply to INADVERTENT CONTAINMENT ISOLATION: Loss of normal heat sink. CFR: 41.8 to 41.10 Proposed Question: RO Question # 24 Given the following:

The plant was operating at 100% power

A maintenance technician performing a test procedure inadvertently initiated a PCIS Group 1 isolation

The resultant transient caused Safety-Relief Valves (SRV) to lift on high Reactor Pressure Vessel (RPV) pressure

After the SRVs close, the operators take NO further action At what pressure will the next Safety-Relief Valve open, and what is the associated basis?

A. 1030 psig; to protect the SRV tailpipes from damage B. 1035 psig; to prevent reaching the high pressure scram setpoint C. 1110 psig; to minimize cycling of the other SRVs D. 1120 psig; to protect the RPV from over-pressurization Proposed Answer: A Explanation (Optional):

A. Correct: a scram signal would be initiated from the MSIV closure, but pressure will rise rapidly due to the steam being bottled up and a high pressure scram signal will be generated at 1055 psig. Pressure will continue to rise since there has been no release of energy and all six of the Safety/Relief Valves will open. The LLS valve logic is now armed and will remain armed until manually reset. Since reactor pressure will be greater than the open setpoint for the LLS valves (the lowest SRV actuation pressure is 1110 psig and the open setpoints for the LLS valves are 1030 psig and 1035 psig) the LLS valves will get an open signal. As pressure comes down below the reset setpoint, the ADS SRVs will shut, but the LLS valves will remain open until they reach their shut setpoint of 915 psig for the high valve and 910 psig for the low valve. At 910 psig all SRVs will close and pressure will begin to rise. When pressure reaches 1030 psig, the low LLS valve will open to control pressure and lower the vessel pressure back down to 910 psig. This will continue until equilibrium is reached or some other method of pressure control such as RCIC or HPCI is placed in service (SD-183.1 pages 12-13).

Also, the purpose of the Low-Low Set System is to mitigate the induced high frequency 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 49

loads on the containment and thrust loads on the SRV discharge lines (SD-183.1 page 4).

B. Incorrect: the high pressure scram setpoint is 1055 psig, but a scram would have already occurred due to MSIV closure (SD-183.1 pages 12-13).

C. Incorrect: the low LLS valve setpoint is 1030 psig. 1110 psig is the SRV setpoint. (SD-183.1 pages 12-13)

D. Incorrect: the low LLS valve setpoint is 1030 psig. 1110 psig is the SRV setpoint. (SD-183.1 pages 12-13).

Technical Reference(s): SD-183.1, Revision 7 Proposed References to be provided to applicants during examination: N Learning Objective: 8.00.00.03 (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments: An I&C tech performing an STP inadvertently initiated a Group 1 isolation while at full power. The resultant transient lifted four Safety-Relief Valves, including the two low-low set valves, on high RPV pressure.

After these valves close, as indicated by their associated amber lights extinguishing, the operators take no further action.

At WHAT PRESSURE will the next Safety-Relief Valve open and WHY?

A. 1020 psig; to protect the SRV tailpipes from damage.

B. 1020 psig; to prevent reaching the high pressure scram setpoint.

C. 1110 psig; to minimize cycling of the other SRVs.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 50

D. 1110 psig; to protect the reactor from overpressurization.

Answer: A 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 51

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295032 A1.01 Importance Rating Ability to operate and/or monitor the following as they apply to HIGH SECONDARY CONTAINMENT AREA TEMPERATURE : Area temperature monitoring system. CFR: 41.7 Proposed Question: RO Question # 25 Given the following:

Indications of a pipe break in the drywell and rising Suppression pool area ambient temperatures at 1C21 are observed

RCIC and HPCI pumps are operating and supplying water to the RPV

Annunciator 1C04B (B-4), Steam leak Det Ambient HI Temp, IS in alarm

Suppression Pool area ambient temperatures reached 151°F twenty (20) minutes ago Assuming no operator action, which of the following describes the expected response of the RCIC and HPCI turbines?

A. Both HPCI and RCIC turbines will have isolated B. Only the HPCI turbine will have isolated C. Only the RCIC turbine will have isolated D. Neither the HPCI nor RCIC turbines will have isolated Proposed Answer: B Explanation (Optional):

A. Incorrect: Suppression Pool ambient temperature instruments will supply inputs to both HPCI and RCIC isolation logic circuits at 150F. Both pumps have isolation circuitry with timers that will begin counting down when this input is received. HPCI and RCIC have different time delays however; 15 min for HPCI and 30 min for RCIC. At the 15 minute point on the HPCI pump would trip (SD-858 pages 9-10).

B. Correct. Suppression Pool ambient temperature instruments will supply inputs to both HPCI and RCIC isolation logic circuits at 150F. Based upon timer settings, HPCI isolation (which causes a pump trip) will occur after 15 minutes (SD-858 pages 9-10).

C. Incorrect: Suppression Pool ambient temperature instruments will supply inputs to both HPCI and RCIC isolation logic circuits at 150F. Based upon timer settings, HPCI isolation (which causes a pump trip) will occur after 15 minutes. RCIC isolation will occur at the 30 minute point (SD-858 pages 9-10).

D. Incorrect: Suppression Pool ambient temperature instruments will supply inputs to both HPCI and RCIC isolation logic circuits at 150F. Both pumps have isolation features (that result in pump trips) which are set to time delay. The time delays are 15 min for 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 52

HPCI and 30 min for RCIC (SD-858 pages 9-10).

Technical Reference(s): ARP 1C04B, Revision 79 SD-150, Revision 8 SD-152, Revision 13 SD-858, Revision 7 Proposed References to be provided to applicants during examination: N Learning Objective: 50007.05.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 53

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295034 EK3.05 Importance Rating Knowledge of the reasons for the following responses as they apply to SECONDARY CONTAINMENT VENTILATION HIGH RADIATION: Manual SCRAM and depressurization:

Plant-Specific Proposed Question: RO Question # 26 A plant event resulted in a steam leak into secondary containment and rising secondary containment ventilation radiation levels and release rates. The Control Room Supervisor has entered EOP 3, Secondary Containment Control.

What purpose does the reactor scram and emergency depressurization achieve in EOP 3?

A. It allows establishment of adequate core cooling using low pressure ECCS pumps B. It reduces the energy in the RPV before reaching conditions where the primary containment will not accommodate an SRV opening C. It places the primary system in a low energy condition to reduce the driving head of the leak D. It places the RPV in a low energy condition before reaching conditions where a loss of coolant accident could not be adequately quenched in the primary containment Proposed Answer: C Explanation (Optional):

A. Incorrect - This is not the purpose of ED for this event. This would be correct in the event of a LOCA and lowering level.

B. Incorrect - Containment parameters such as increasing drywell pressure are not an issue in the described event. Accommodating SRV openings is not an issue for this event C. Correct - Scramming the reactor reduces the energy that the RPV may be discharging to the secondary containment to decay heat levels. If the RPV is the source of energy, radiation or water being released to secondary containment, scramming the reactor should greatly reduce any further release and may prevent the need for the more severe action of emergency depressurizing the RPV. RPV depressurization places the primary system in its lowest possible energy state, rejects heat to the torus in preference to outside the containment, and reduces the driving head and flow of primary systems that are unisolated and discharging into the secondary containment.

D. Incorrect - The concern is not a LOCA in EOP 3.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 54

Technical Reference(s): Bases-EOP 3, Rev. 11 Proposed References to be provided to applicants during examination: N Learning Objective: 95.00.00.03 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 55

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 500000 EK 3.01 Importance Rating Knowledge of the reasons for the following responses as they apply to HIGH PRIMARY CONTAINMENT HYDROGEN CONCENTRATIONS: Initiation of containment atmosphere control system. CFR: 41.5 Proposed Question: RO Question # 27 After an event, you are directed to place the containment hydrogen analyzers in service.

Current plant conditions are as follows:

Drywell pressure is 3.2 psig and stable

RPV level is 100 inches and stable

Drywell temperature is 175°F and stable

Hydrogen Analyzers have remained in standby throughout the event

Reactor Building Vent Radiation Monitors are reading 30 mrem and stable To perform this action you must?

A. Install Defeat 9, Group 3 High DW Press & RX Low Level Isolation Defeat, and place the analyzers in service at 1C-09 B. Install Defeat 16, Containment Atmosphere Monitoring Sample Line Isolation Defeat, and place the analyzers in service at 1C-09 C. Install Defeat 9, Group 3 High DW Press & RX Low Level Isolation Defeat, the analyzers will need a 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months warm-up time until the reading is valid D. Install Defeat 16, Containment Atmosphere Monitoring Sample Line Isolation Defeat, the analyzers will need a 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months warm-up time until the reading is valid Proposed Answer: B Explanation (Optional):

A. Incorrect - EOP 3 directs Defeat 9 to be used to re-establish main plant ventilation for reactor building cooling.

B. Correct C. Incorrect - OI 873, Containment Atmosphere Monitoring System, contains a note stating that normally both H2-O2 Analyzers are left in standby to avoid the recommended 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months warm-up of the sample chamber if the analyzer was off. From the conditions provided in the STEM, power was never lost to the analyzers and a warm-up is not required.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 56

D. Incorrect - OI 873, Containment Atmosphere Monitoring System, contains a note stating that normally both H2-O2 Analyzers are left in standby to avoid the recommended 6 hour6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months warm-up of the sample chamber if the analyzer was off. From the conditions provided in the STEM, power was never lost to the analyzers and a warm-up is not required.

Technical Reference(s): EOP Defeat 9, Rev. 4 EOP Defeat 16, Rev. 4 OI 873, Rev. 54 Proposed References to be provided to applicants during examination: N Learning Objective: 95.26.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 57

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 203000 K4.13 Importance Rating (RHR/LPCI Injection Mode) Knowledge of RHR/LPCI: INJECTION MODE (PLANT SPECIFIC) design feature(s) and/or interlocks which provide for the following: Adequate pump net positive suction head (interlock suction valve open): Plant-Specific CFR: 41.7 Proposed Question: RO Question # 28 With the A Loop of Residual Heat Removal (RHR) in Shutdown Cooling, the following annunciators were received:

1C03B(A-5), LPCI HI Drywell Press

1C03B(A-2), A RHR Pump 1P-229A Trip

1C03B(A-3), C RHR Pump 1P-229C Trip

1C03B(D-4), A/B RHR HX RHR Inlet HI TEMP At the time, the following conditions existed:

Drywell pressure was 3 psig

RHR Heat Exchanger inlet temperature was 350°F Which of the following was the reason for the RHR pump trips? To prevent.

A. water hammer to piping on pump auto start B. pump damage due to cavitation C. overpressurization of the low pressure Shutdown Cooling piping D. thermal shock to the vessel due to cold water injection on pump auto start Proposed Answer: B Explanation (Optional):

A. Incorrect: RHR Pump breakers will not close in due to the suction path interlock.

B. Correct: Caused by a loss of NPSH.

C. Incorrect: The Group 4 Isolation signal (135 psig) does this function.

D. Incorrect: RHR Pump breakers will not close in due to the suction path interlock AND RHR pump Suction valves do not automatically open.

Technical Reference(s): SD-149, Rev. 13 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 58

Proposed References to be provided to applicants during examination: N Learning Objective: 2.11.01.14 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2001 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

With the A Loop of RHR in Shutdown Cooling, the following annunciators were received:

1C03B, A-5, LPCI HI DRYWELL PRESS then:

1C03B, A-2, A RHR PUMP 1P-229A TRIP

1C03B, A-3, C RHR PUMP 1P-229C TRIP

1C03B, D-4, A/B RHR HX RHR INLET HI TEMP Drywell pressure is 3 psig RHR Hx inlet temperature is 350°F What was the reason for the RHR pump trips?

a. prevent RHR pump damage due to cavitation caused by a loss of NPSH

b. prevent water hammer to piping when the RHR pumps try to auto start

c. prevent overpressurization of the low pressure Shutdown Cooling piping

d. prevent a thermal shock to the vessel due to injection of cold water from the torus when the RHR pumps auto start ANSWER: a Distracter 1: RHR Pump breakers will not close in due to the suction path interlock.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 59

Distracter 2: The Group 4 Isol signal 135 psig does this function.

Distracter 3: RHR Pump breakers will not close in due to the suction path interlock.

RHR pump Suction valves do not automatically open 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 60

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 205000 K5.03 Importance Rating (Shutdown Cooling) Knowledge of the operational implications of the following concepts as they apply to SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE): Heat removal mechanisms. CFR: 41.5 Proposed Question: RO Question # 29 Given the following:

The plant is in MODE 4 with irradiated fuel in the vessel

Shutdown Cooling (SDC) flow is 4500 gpm

Reactor water level is 235 inches, as indicated on LI-4541, WR GEMAC, FLOODUP What action are required, if any, and why?

A. No actions are required for the given plant conditions B. Raise Reactor water level; it is below the minimum natural circulation level C. Raise Reactor water level; it is below the minimum SDC NPSH requirement D. Lower Shutdown cooling flow, to prevent lifting the steam dryer Proposed Answer: A Explanation (Optional):

A. Correct: All parameters provided in the STEM are within allowable limits.

B. Incorrect. The minimum natural circulation level is 214 inches.

C. Incorrect. This is water level is well above the SDC NPSH requirements.

D. Incorrect. OI 149. P&L 20 on page 7, states that the maximum shutdown cooling flow is

<4800 gpm to ensure that the recirc pump is not dead headed.

Technical Reference(s): OI 149, Rev.

Proposed References to be provided to applicants during examination: N 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 61

Learning Objective: 2.11.01.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 62

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 205000 K1.01 Importance Rating 3.6 Knowledge of the physical connections and/or cause- effect relationships between SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE) and the following:

Reactor pressure.

Proposed Question: RO Question # 30 Following a reactor scram, the following conditions exist:

Reactor level +190 inches

Reactor pressure 139 psig

Drywell pressure 1.72 psig Based upon the given conditions, which ONE of the following Residual Heat Removal valves is interlocked closed/prevented from opening?

A. MO-2006, RHR Loop A Torus Spray Header Isolation Valve B. MO-1908, RHR Shutdown Cooling Isolation Valve C. MO-2007, RHR Loop A Torus Cooling and Test Return HDR Isolation Valve D. MO-1940, RHR HX 1E-201B Bypass Valve Proposed Answer: B Explanation (Optional):

A. Incorrect - For the given conditions MO-2006 is able to be opened, the valve is isolated when containment pressure is > 2 psig. This is plausible because the candidate may assume that a high drywell pressure is needed to place torus sprays in service.

B. Correct - Of the signals listed; only the reactor pressure signal causes an RHR isolation/interlock. This high-pressure interlock prevents the SDC section of piping from being over pressurized. A reactor pressure of approximately 135 psig (per ARP 1C03B B-4 this pressure is approximately 100 psig) initiates an isolation of SDC suction valves MO-1908 and 1909. The LPCI piping is also protected from over pressurization, but the setpoint is 450 psig.

C. Incorrect - For the given conditions MO-2007 is able to be opened, the valve is isolated when containment pressure is > 2 psig. This is plausible because the candidate may assume that a high drywell pressure is needed to place torus sprays in service.

D. Incorrect - MO-1940 has an automatic open function on a LPCI initiation signal. It does NOT have an auto close function.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 63

Technical Reference(s): ARP 1C05B, D-8, Rev. SD-149, Rev. 13, pages. 31-34 Proposed References to be provided to applicants during examination: N Learning Objective: 2.11.01.14 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2013 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 64

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 206000 K6.12 Importance Rating Knowledge of the effect that a loss or malfunction of the following will have on the HIGH PRESSURE COOLANT INJECTION SYSTEM: Reactor water level. CFR: 41.7 Proposed Question: RO Question # 31 Given the following:

The plant was operating at 100% power when a loss of coolant accident occurred

Reactor Pressure Vessel (RPV) Pressure is currently 400 psig

RHR and Core Spray pumps are running

RPV level is currently +20 inches and continues to lower Based on the conditions above, which of the following actions should operators take with regard to:

(1) Emergency Depressurization (ED)

AND (2) What is the status of the High Pressure Coolant Injection (HPCI) system?

A. (1) ED immediately (2) HPCI is insufficient to maintain RX water level B. (1) ED when RPV level is +15 to -25 inches (2) HPCI is insufficient to maintain RX water level C. (1) ED once RPV level is below -25 inches (2) HPCI is insufficient to maintain RX water level D. (1) ED immediately (2) HPCI should be tripped prior to the low pressure isolation setpoint Proposed Answer: B Explanation (Optional):

A. Incorrect: If an injection source is available, the blowdown should be delayed at least until RPV water level reaches the top of the active fuel (+15 in). If it is believed that available injection systems are capable of restoring and maintaining RPV water level above the Minimum Steam Cooling RPV Water Level following RPV depressurization, the blowdown may be performed as soon as RPV water level reaches the top of the 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 65

active fuel. (EOP 1 Bases, pages 36 -37)

B. Correct: If an injection source is available, the blowdown should be delayed at least until RPV water level reaches the top of the active fuel (+15 in), but may be performed anytime RPV water level is between the top of the active fuel and the Minimum Steam Cooling RPV Water Level (-25 in). If it is believed that available injection systems are capable of restoring and maintaining RPV water level above the Minimum Steam Cooling RPV Water Level following RPV depressurization, the blowdown may be performed as soon as RPV water level reaches the top of the active fuel. (EOP 1 Bases, pages 36 -37)

C. Incorrect: The core will remain adequately cooled as long as RPV water level remains above the MSCRWL (-25 in.) The MSCRWL (-25 in.) is the lowest RPV water level at which the covered portion of the reactor core will generate sufficient steam to preclude any clad temperature in the uncovered portion of the core from exceeding 1500°F.

(EOP 1 Bases, pages 36 -37)

D. Incorrect: the HPCI system would have needed to fail to automatically start at the low-low RPV water level setpoint of 119.5 inches for current RPV water to level to 20 inches and lowering with RPV pressure still at 400 psig. It must be recognized that HPCI has failed and that now emergency depressurization will be required when the appropriate RPV level is reached (SD-152 pages 18-22). The candidate may incorrectly assume that operator action is required to trip and isolate the HPCI turbine prior to the lower pressure isolation setpoint.

Technical Reference(s): EOP 1, Revision 18 Bases-EOP 1, Revision 16 SD-152, Revision 13 SD-183.1, Revision 7 Proposed References to be provided to applicants during examination: N Learning Objective: 95.74.12.01/5.06.01.04 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 66

Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 67

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 206000 A3.01 Importance Rating Ability to monitor automatic operations of the HIGH PRESSURE COOLANT INJECTION SYSTEM including: Turbine speed. CFR: 41.7 Proposed Question: RO Question # 32 Given the following:

The plant was operating at 50% power when rising drywell pressure resulted in a scram

1C03C(A-3), HPCI AUTO Initiated, was lit and flow to the RPV from HPCI was observed

The following HPCI parameters were observed:

o HPCI steam line pressure was 900 psig o HPCI steam line flow was 125%

o HPCI turbine exhaust pressure was 100 psig Subsequently, the following annunciators were observed:

1C03C(A-4), HPCI Turbine Tripped, IS in alarm

1C03C(A-5), HPCI Turbine Trip Solenoid Energized, was NOT in alarm Which of the following identifies the cause of the HPCI turbine response?

A. Turbine Exhaust Pressure High B. Steam Line Flow High C. Steam Line Pressure Low D. Turbine Overspeed Proposed Answer: D Explanation (Optional):

A. Incorrect: High turbine exhaust pressure would energize the turbine trip solenoid and cause ARP 1C03C (A 5), HPCI TURBINE TRIP SOLENOID ENERGIZED, to illuminate.

Also, turbine exhaust pressure is provided in the stem at 100 psig, but the trip does not occur until 140 psig. (SD-152 pages 20-27 and 1C03C A-5/B-5)

B. Incorrect: High HPCI Steamline Flow will result in a HPCI Isolation signal from both HPCI Isolation logic trains. This in turn will cause a trip via the turbine trip solenoid.

Also, steam line flow is given in the stem at 125%, but the trip does not occur until 300%. (SD-152 pages 20-27 and 1C03C A-4/B-8) 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 68

C. Incorrect: Low HPCI Steam Line Pressure would energize the turbine trip solenoid and cause ARP 1C03C (A 5), HPCI TURBINE TRIP SOLENOID ENERGIZED, to illuminate.

Also steamline pressure is given in the stem as 900 psig, but the trip occurs at a setpoint of 50 psig < P < 100psig. (SD-152 pages 20-27 and 1C03C A-5)

D. Correct: 1C03C (A 4), HPCI TURBINE TRIPPED, contains a note informing which states that ARP 1C03C (A 5), HPCI TURBINE TRIP SOLENOID ENERGIZED, will remain clear if the trip results from turbine mechanical overspeed. As described in SD-152 (pages 22-24), the HPCI turbine overspeed trip operates via a mechanical mechanism that is independent of the solenoid trip functions.

Technical Reference(s): ARP 1C03C, Revision 41 OI-152, Revision 110 SD-152, Revision 13 Proposed References to be provided to applicants during examination: N Learning Objective: 5.06.01.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 69

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 203000 K2.02 Importance Rating Knowledge of electrical power supplies to the following: Valve power. CFR: 41.2 to 41.9 Proposed Question: RO Question # 33 Given the following:

The plant is operating at 100% reactor power

Bus 1A3 suffered a lockout trip NOTE: Table provided to identify LPCI inject valves and recirc valves being evaluated.

A LPCI and Recirc Valves B LPCI and Recirc Valves MO-2003 INBD Inject Isolation MO-1905 INBD Inject Isolation MO-2004 OUTBD Inject Isolation MO-1904 OUTBD Inject Isolation MO-4601 Recirc Pump Inlet MO-4602 Recirc Pump Inlet MO-4627 Discharge Isolation MO-4628 Discharge Isolation MO-4629 Discharge Bypass MO-4630 Discharge Bypass What is the status of the following equipment?

A. A LPCI valves are ENERGIZED, A recirc valves are ENERGIZED B LPCI valves are ENERGIZED, B recirc valves are ENERGIZED B. A LPCI valves are ENERGIZED, A recirc valves are DE-ENERGIZED B LPCI valves are ENERGIZED, B recirc valves are ENERGIZED C. A LPCI valves are ENERGIZED, A recirc valves are DE-ENERGIZED B LPCI valves are ENERGIZED, B recirc valves are DE-ENERGIZED D. A LPCI valves are DE-ENERGIZED, A recirc valves are DE-ENERGIZED B LPCI valves are ENERGIZED, B recirc valves are ENERGIZED Proposed Answer: A Explanation (Optional):

A. Correct - SD-304, page 34 states that [1B33A/1B34A] supply power to the RHR injection valves MO-1905, MO-1904, MO-2003, and MO-2004, Recirculation Pump Suction Valves MO-4601, MO4602, Recirculation Pump Discharge Valves MO-4627, MO4628, and Recirculation Pump Discharge Bypass Valves MO-4629, MO-4630. The automatic transfer switch assures that power will be available for valve operation.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 70

B. Incorrect - A recirc valves will be energized.

C. Incorrect - A recirc valves will be energized.

D. Incorrect - See correct answer description.

Technical Reference(s): SD-304, Revision 19 AOP 301, Rev. 65 SD-159, Rev. 13 Proposed References to be provided to applicants during examination: N Learning Objective: 2.03.01.20 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 71

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 209001 A2.08 Importance Rating Ability to (a) predict the impacts of the following on the LOW PRESSURE CORE SPRAY SYSTEM; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve openings malfunctions.

CFR 41.5 Proposed Question: RO Question # 34 The plant was operating at 100% when a recirc line break occurred. Current plant conditions are:

Reactor pressure is at 410 psig and stable

Reactor level is at 60 inches and rising slowly

Drywell Pressure is at 3.4 psig and rising slowly

Core Spray pumps are running

Core Spray Inboard Injection Valves, MO-2117 and MO-2137, are CLOSED

Core Spray Minimum Flow Bypass Valves, MO-2104 and MO-2124, are OPEN Which of the following describes the response of the Core Spray System to the current plant conditions?

Core Spray inboard inject valves __(1)__.

AND Core Spray minimum flow bypass valves __(2)__.

A. (1) should have opened and must be manually opened (2) will auto-close ONLY when the Injection Valves are fully open B. (1) are closed and will open once reactor pressure lowers to below the shut off head of the Core Spray pumps, no operator action is required (2) will auto-close when Core Spray system flow reaches 600 gpm C. (1) should have opened and must be manually opened (2) will auto-close when Core Spray system flow reaches 600 gpm D. (1) are closed and will open once reactor pressure lowers to below the shut off head of the Core Spray pumps, no operator action is required (2) will auto-close ONLY when the Injection Valves are fully open Proposed Answer: C 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 72

Explanation (Optional):

A. Incorrect - The min flow bypass valve will close when system flow reaches 600 gpm.

B. Incorrect - The injection valves should have opened at 450 psig RPV pressure and must be opened manually based on the step stating to verify they open.

C. Correct - OI 151, pages 6 and 7, steps 4.0 (2) and (3). When system flow reaches 600 gpm, as indicated on (A[B] CORE SPRAY PUMP) INJECT/TEST FLOW indicator FI-2110 [FI-2130] on Panel 1C03, verify MIN FLOW BYPASS MO-2104 [MO-2124] valve CLOSES. When reactor vessel pressure drops below the low pressure permissive setpoint of 450 psig, verify that the INBD INJECT MO-2117 [MO-2137] valves OPEN to inject to the reactor vessel. The injection valves should have opened at 450 psig RPV pressure and must be opened manually based on the step stating to verify they open.

D. Incorrect - The injection valves should have opened at 450 psig RPV pressure and must be opened manually based on the step stating to verify they open. The min flow bypass valve will close when system flow reaches 600 gpm.

Technical Reference(s): OI 151, Rev 74 Proposed References to be provided to applicants during examination: N Learning Objective: 4.02.01.07 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 73

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 211000 A4.01 Importance Rating Ability to manually operate and/or monitor in the control room: Tank level. CFR: 41.7 Proposed Question: RO Question # 35 Given the following:

The plant was operating at 100% when an Anticipated Transient Without Scram (ATWS) occurred

Standby Liquid Control (SBLC) was initiated, and both SBLC pumps were verified to be operating and delivering their design flow

Twenty (20) minutes after SBLC initiation, tank level is checked and observed to be zero (0) on Control Room Panel 1C05 Based upon these conditions, the SBLC tank level ____(1)____ and Cold Shutdown Boron Weight ____(2)____ been injected into the RPV.

A. (1) is empty (2) has B. (1) is empty (2) has NOT C. (1) indication has failed (2) has D. (1) indication has failed (2) has NOT Proposed Answer: D Explanation (Optional):

A. Incorrect: each SBLC pump is designed to supply 26.2 gpm. Both pumps are operated by a common switch and running together would deliver approximately 52.4 gpm. After 35 minutes, this would have introduced a maximum of 1834 gallons of boron to the RPV. The tank would not yet indicate zero based on actual inventory (when the low level alarm actuates, 2600 gallons remain in the tank; this alarm is not provided as illuminated in the initial conditions). Additionally, the required available quantity of boron is a function of concentration, but is always greater than 2000 gallons. Thus adequate boron could not yet have been added in this case regardless of concentration, temperature, etc. (SD-153 pages 6 -15).

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 74

B. Incorrect: each SBLC pump is designed to supply 26.2 gpm. Both pumps are operated by a common switch and running together would deliver approximately 52.4 gpm. After 35 minutes, this would have introduced a maximum of 1834 gallons of boron to the RPV. The tank would not yet indicate zero based on actual inventory (when the low level alarm actuates, 2600 gallons remain in the tank; this alarm is not provided as illuminated in the initial conditions). (SD-153 pages 6 -15)

C. Incorrect: level indication for the SBLC tank is supplied by a bubbler which is in turn supplied by instrument air. A failure of the air supply to this bubbler would cause the indicator to fail low regardless of actual tank level (SD-153 page 8 and AOP-518 page 9). Each SBLC pump is designed to supply 26.2 gpm. Both pumps are operated by a common switch and running together would deliver approximately 52.4 gpm. After 35 minutes, this would have introduced a maximum of 1834 gallons of boron to the RPV.

The required available quantity of boron is a function of concentration, but is always greater than 2000 gallons. Thus adequate boron could not yet have been added in this case regardless of concentration, temperature, etc.(SD-153 pages 6 -15).

D. Correct: level indication for the SBLC tank is supplied by a bubbler which is in turn supplied by instrument air. A failure of the air supply to this bubbler would cause the indicator to fail low regardless of actual tank level. (SD-153 page 8 and AOP-518 page 9)

Technical Reference(s): ARP 1C05A, Revision 78 AOP 518, Revision 34 SD-153, Revision 8 Proposed References to be provided to applicants during examination: N Learning Objective: 6.00.00.04 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 75

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A # 212000 RPS, 2.4.2: Knowledge of how abnormal operating procedures are used in conjunction with EOPs. CFR: 41.10 Proposed Question: RO Question # 36 Given the following:

The plant is operating at 100% reactor power

RPV water level is 190 inches and stable

HPCI starts due to a VALID signal Which ONE of the following describes the effect of this VALID signal on the plant and the required procedure entry?

A. The reactor will scram when RPV Water Level rises to 211 inches, which requires entry into IPOI-5, Reactor Scram ONLY B. The reactor will immediately scram; it is required to enter EOP-1, RPV Control AND EOP-2, Primary Containment Control C. The reactor will immediately scram; it is required to enter EOP-1, RPV Control ONLY D. The reactor will scram when APRM Power Level rises, which requires entry into EOP-1, RPV Control, AND IPOI-5, Reactor Scram Proposed Answer: B Explanation (Optional):

A. Incorrect. Plausible; would be true for a Manual HPCI Start resulting in overfeed.

B. Correct. HPCI Auto Start with RPV Water Level at 195 inches implies Drywell Pressure

> 2.0 psig. The reactor will immediately scram, EOP-1 and EOP-2 entry are required with Drywell Pressure > 2.0 psig.

C. Incorrect. Plausible: would be true for Low RPV Water Level (119.5 inches) HPCI start signal, excluded by RPV Water Level at 195 inches.

D. Incorrect. Plausible: would be true for a spurious HPCI start, reactivity addition would cause APRM Power to rise, excluded by valid HPCI start signal.

Technical Reference(s): ARP 1C05B A-1, Rev. EOP-1, Rev.

EOP-2, Rev.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 76

Proposed References to be provided to applicants during examination: N Learning Objective: 22.00.00.03 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 77

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 215003 K1.04 Importance Rating Knowledge of the physical connections and/or cause effect relationships between INTERMEDIATE RANGE MONITOR (IRM) SYSTEM and the following: Process computer /

performance monitoring system (SPDS/ERIS/CRIDS/GDS). CFR: 41.2 to 41.9 Proposed Question: RO Question # 37 Given the following:

The plant is in MODE 2

The "B" Intermediate Range Monitor (IRM) is placed in BYPASS With the B IRM bypassed, which SET of "B" IRM indications are CORRECTLY described below?

1 - ALL "B" IRM 1C05 indicating lamps on the Reactor Control Benchboard are defeated 2 - ALL "B" IRM outputs to the recorder are defeated (NOTE: word "to" was changed from "from" in this item by NRC based on a question by an applicant during the exam) 3 - ALL "B" IRM outputs to the annunciators are defeated 4 - ALL "B" IRM channel inputs to SPDS remain available 5 - ALL 1C36 indications for the "B" IRM are defeated A. 1, 3, 5 B. 2, 3, 4 C. 1, 2, 4 D. 2, 3, 5 Proposed Answer: B Explanation (Optional):

A. Incorrect - The Retract Permit Lamp will remain LIT on 1C05 as long as the IRM channel is bypassed and the IRM detector is not full out. ALSO, Panel 1C-36 has an IRM BYPASSED light for each of the six IRM channels.

B. Correct - When an IRM channel is bypassed, the following IRM functions are defeated:

a. The IRM UPSCALE trip to Reactor Protection System. b. The IRM associated trips to the rod withdrawal block circuits of the Reactor Manual Control System. c. The IRM outputs to the annunciator and sequence recorder. d. The IRM outputs to the indicating lamps on the Reactor Control Benchboard. The Retract Permit Lamp will remain ON as long as the IRM channel is bypassed and the IRM detector is not full out.

C. Incorrect - The Retract Permit Lamp will remain LIT on 1C05 as long as the IRM channel is bypassed and the IRM detector is not full out 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 78

D. Incorrect - Panel 1C-36 has an IRM BYPASSED light for each of the six IRM channels.

Technical Reference(s): OI-878.2, Rev 24 SD 878.2, Rev 9 Proposed References to be provided to applicants during examination: N Learning Objective: 79.01.01.01 (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam: 2011 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 79

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 215004 K2.01 Importance Rating Knowledge of electrical power supplies to the following: SRM channels/detectors. CFR: 41.7 Proposed Question: RO Question # 38 Given the following:

The plant is in MODE 2

Source Range Monitors (SRM) are completely inserted into the core

Power is lost to 1D60, 24 VDC Distribution Panel Div 2 Based on the conditions above, which of the following SRMs would be expected to have lost power?

A. SRMs A AND B B. SRMs A AND C C. SRMs B AND D D. SRMs C AND D Proposed Answer: C Explanation (Optional):

A. Incorrect: the failure of 1D60 (24 VDC Division II) will cause a loss of power to SRMs B and D. SRMs A and C are powered off of 1D50 (24 VDC Division I) and would still have power. (SD-878.1 page 28)

B. Incorrect: the failure of 1D60 (24 VDC Division II) will cause a loss of power to SRMs B and D. SRMs A and C are powered off of 1D50 (24 VDC Division I) and would still have power. (SD-878.1 page 28)

C. Correct.

D. Incorrect: the failure of 1D60 (24 VDC Division II) will cause a loss of power to SRMs B and D. SRMs A and C are powered off of 1D50 (24 VDC Division I) and would still have power. (SD-878.1 page 28)

Technical Reference(s): SD-375, Revision 8 SD-878.1, Revision 7 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 80

Proposed References to be provided to applicants during examination: N Learning Objective: 23.00.00.03 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 81

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 215004 K5.03 Importance Rating Knowledge of the operational implications of the following concepts as they apply to SOURCE RANGE MONITOR (SRM) SYSTEM: Changing detector position. CFR: 41.5 Proposed Question: RO Question # 39 Given the following:

A reactor startup is in progress

All Intermediate Range Monitor (IRM) range switches are selected to range 3 or 4

Source Range Monitors (SRMs) are still fully inserted into the core

The SRM UPSCALE setpoint has just been exceeded Which of the following describes the operational effects that would be expected to result from this condition?

A. An SRM UPSCALE rod block occurs and the operator cannot insert or withdraw control rods B. An SRM UPSCALE annunciator illuminates but a rod block does NOT occur C. An SRM UPSCALE trip occurs and a reactor scram results D. An SRM UPSCALE rod block occurs and the operator cannot withdraw control rods Proposed Answer: D Explanation (Optional):

A. Incorrect: a rod block does occur at 1x105 CPS, however rod insertion is not blocked.

The rod block circuitry is configured such that the relays associated with the rod block addressed in this question affect the signal path for rod withdrawal, however still leave the rod insertion path available. (SD-856 pages 14-17)

B. Incorrect: a rod block does occur as a result of the SRM upscale at 1x105 CPS.

Additionally, the required IRM range would be 7, not 3, to affect this block (range 3 would be affect a downscale condition). (SD-878 pages 24-25)

C. Incorrect: a scram would occur at 5x105 CPS in the event that shorting links were not connected across the initial fuel loading relay contacts. It must be understood that the significance of the shorting links being installed is that the RPS input is thereby blocked, and not enabled. (SD-878 page 28)

D. Correct: If the SRM detectors are not retracted, the SRM UPSCALE rod block setpoint may be exceeded. This will cause a rod withdrawal block with its attendant alarms. (OI-878 page 6) 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 82

Technical Reference(s): OI-878.1, Revision 19 SD-856.1, Revision 8 SD-878.1, Revision 7 Proposed References to be provided to applicants during examination: N Learning Objective: 82.00.00.04 (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments: Original DAEC Bank Question:

A reactor startup is in progress. While performing the startup, the RO fails to retract the SRM detectors before the SRM UPSCALE setpoint is exceeded. All IRM range switches are on range 3 or 4.

What automatic actions occur, if any, due to this condition?

A. An SRM UPSCALE rod block occurs, and the operator cannot withdraw control rods until SRM counts are below the reset point.

B. An SRM UPSCALE rod block occurs, and the operator cannot withdraw or insert control rods until SRM counts are below the reset point.

C. An SRM UPSCALE annunciator illuminates and warns the operator that the SRM counts are high. No rod block occurs because all IRMs are on range 3 or above.

D. An SRM UPSCALE trip will cause a scram if one SRM from each RPS channel reaches its upscale trip setpoint.

Answer: A 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 83

Answer Explanation:

50007_878.1_lp rev 1 OI-878.1 Rev. 19, page 6, CAUTION statement If the SRM detectors are not retracted, the SRM UPSCALE rod block setpoint (105 cps) may be exceeded. This will cause a rod withdrawal block with its attendant alarms. A rod withdrawal block will also occur if the SRM detectors are retracted to the point where the flux level is lower than the detector Retract Permissive setpoint (100 cps).

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 84

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 215005 K 3.07 Importance Rating Knowledge of the effect that a loss or malfunction of the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM will have on following: Rod block monitor. (CFR: 41.7 / 45.4)

Proposed Question: RO Question # 40 Give the following:

The plant is operating at 78% reactor power

A NON-edge control rod is selected

"C" level LPRM on the Four Rod Display fails DOWNSCALE

Rod Block Monitor is reading 66/125 Which ONE of the following describes the effect of this failure on the Rod Block Monitor System?

This failure affects the input to .

A. BOTH Rod Block Monitors and WILL result in an RBM Rod Block B. ONLY ONE Rod Block Monitor and WILL result in an RBM Rod Block C. ONLY ONE Rod Block Monitor and WILL NOT result in an RBM Rod Block D. BOTH Rod Block Monitors and WILL NOT result in an RBM Rod Block Proposed Answer: A Explanation (Optional):

A. Correct - C level LPRMs are inputs to both RBMs which will produce RBM Rod Blocks when indication lowers below 94/125.

B. Incorrect - would be true for a B or D level LPRM.

C. Incorrect - would be true for a B or D level LPRM below 30% initial power.

D. Incorrect - would be true below 30% initial power.

Technical Reference(s): SD-878.5, Rev 10, p. 10/15 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 85

Proposed References to be provided to applicants during examination: N Learning Objective: 82.00.00.04 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

[K/A 215002 K 6.05 Knowledge of the effect that a loss or malfunction of the following will have on the ROD BLOCK MONITOR SYSTEM : LPRM detectors (CFR: 41.7 / 45.7)]

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 86

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 215005 A1.04 Importance Rating Ability to predict and/or monitor changes in parameters associated with operating the AVERAGE POWER RANGE MONITOR/LOCAL POWER RANGE MONITOR SYSTEM controls including: SCRAM and rod block trip setpoints. CFR: 41.5 Proposed Question: RO Question # 41 Given the following:

Reactor power has just been raised to 62%

Both recirculation loops are in operation, with flow at 67% per loop

Total Core Flow is 33 Mlbm/hr (equally divided between both loops)

With the current plant conditions, what would be the Average Power Range Monitor flow biased ROD BLOCK setpoint?

A. 64%

B. 71%

C. 83%

D. 90%

Proposed Answer: D Explanation (Optional):

A. Incorrect: miscalculated using a value for W of 33 based on total core flow (vice using the correct percent recirculation loop flow value of 67) and the single loop equation constant value of 46 (vice the correct two loop constant of 53). (SD-878.3 page 24)

B. Incorrect: miscalculated using a value for W of 33 based on total core flow (vice using the correct percent recirculation loop flow value of 67). (SD-878.3 page 24)

C. Incorrect: miscalculated using the single loop equation constant value of 46 (vice the correct two loop constant of 53). (SD-878.3 page 24)

D. Correct: APRM flow biased rod block occurs with two recirculation loops in operation at 0.55W + 53, where W equals the percent of recirculation flow. The value 53 is a constant that applies when both recirculation loops are in operation. (SD-878.3 page 24)

Technical Reference(s): SD-878.3, Revision 11 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 87

Proposed References to be provided to applicants during examination: N Learning Objective: 82.00.00.04 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 88

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 261000 K4.03 Importance Rating Knowledge of REACTOR CORE ISOLATION COOLING SYSTEM (RCIC) design feature(s) and/or interlocks which provide for the following: Prevents pump overheating. CFR: 41.7 Proposed Question: RO Question # 42 Given the following:

The plant is operating at 100% reactor power

The Balance of Plant Operator is cycling MO-2510, RCIC Pump Minimum Flow Bypass Valve, for post maintenance testing

MO-2510 will be cycled from 1C04 using the valves handswitch What is the expected response of MO-2510 when the valves handswitch is taken to the OPEN position and then immediately returned to the AUTO position?

A. The valve will open and remain open B. The valve cannot be opened by the handswitch C. The valve will go to the fully open position and then automatically close D. The valve cannot be opened unless Defeat 1, RCIC Low RPV Pressure Isolation and 211 Defeat, is installed Proposed Answer: C Explanation (Optional):

A. Incorrect - The valve will fully open. Due to plant conditions provided in the STEM, MO-2510 will automatically close.

B. Incorrect - The valve will fully open with the valve handswitch.

C. Correct - SD-150 provides the MO-2510 RCIC Minimum Flow Valve Logic figure which requires that RCIC discharge pressure to be > 125 psig and a flow < 40 gpm for MO-2510 to remain open. This valve control logic is equipped with a seal in function to prevent redirecting the valve during valve intermediate positioning. A 2 second time delay relay has been installed on the valve control logic which remains in the circuit when the valve handswitch has been returned to AUTO.

D. Incorrect - Defeat 1 is used to maintain the RCIC steam supply available following trip condition of reactor water level at 211 inches. MO-2510, operation does not require this defeat to be installed to operate.

Technical Reference(s): SD-150, Revision 8 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 89

Proposed References to be provided to applicants during examination: N Learning Objective: 3.02.01.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 90

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 218000 K5.01 Importance Rating Knowledge of the operational implications of the following concepts as they apply to AUTOMATIC DEPRESSURIZATION SYSTEM: ADS logic operation. CFR: 41.5 Proposed Question: RO Question # 43 Given the following:

At time 10:00:00, a loss of coolant accident (LOCA) occurred. Standby Diesel Generators 1G-31 and 1G-21 failed to automatically start

At time 10:03:00, 1C03A(A-5), ADS "A/B" 2 MIN Timer(s) Initiated, IS LIT

At time 10:04:00, a loss of offsite power occurred

At time 10:06:00, Standby Diesel Generator 1G-31 was started manually and is powering bus 1A3

It is CURRENTLY 10:06:30. RPV level is 20 inches and has lowered continually since the start of the LOCA Which of the following describes the response of the Automatic Depressurization System (ADS) when the A Core Spray pump started?

A. ADS will initiate immediately B. ADS will initiate 30 seconds later C. ADS will initiate 90 seconds later D. ADS will not initiate automatically Proposed Answer: A Explanation (Optional):

A. Correct: ADS logic is DC powered and thus the ADS time delay relay continued counting down after the site blackout. The two minute countdown would have concluded at time 10:05:00. By 10:06:30, the only remaining input needed for ADS initiation is the start of either an RHR or CS pump (specifically their discharge pressure).

Thus ADS will initiate immediately. (SD-183 pages 14 - 17)

B. Incorrect. If it is not understood that the ADS time delay relay countdown continued during the site blackout, then 30 seconds would appear to remain following the last logic input (RHR or CS pump start occurring). (SD-183 pages 14 - 17)

C. Incorrect: If it is not understood that the ADS time delay logic does not restart from time zero when AC power is restored, then it would appear that 90 seconds must still elapse before ADS initiation. (SD-183 pages 14 - 17) 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 91

D. Incorrect: If it is not understood that the ADS logic circuit has retained its contact states following the time delay relay timing out (which would have occurred 1.5 minutes before time 10:06:30), then it would appear that an automatic ADS initiation will no longer occur since the logic would no longer be satisfied. (SD-183 pages 14 - 17)

Technical Reference(s): SD-183.1, Revision 7 Proposed References to be provided to applicants during examination: N Learning Objective: 8.03.01.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 92

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 223002 K6.05 Importance Rating Knowledge of the effect that a loss or malfunction of the following will have on the PRIMARY CONTAINMENT ISOLATION SYSTEM/NUCLEAR STEAM SUPPLY SHUT-OFF: Containment instrumentation. CFR: 41.7 Proposed Question: RO Question # 44 Given the following:

The plant was operating at 30% when a leak occurred in the drywell

Drywell pressure is currently 2.5 psig and rising slowly

Reactor Pressure Vessel water level lowers to 165 inches and is stable

Due to calibration errors, Primary Containment High Pressure Trip Channels A2 and B2 have both failed to trip Which of the following describes the automatic response of Primary Containment Isolation Group 2 valves to these conditions?

A. neither inboard nor outboard valves will close B. only inboard valves will close C. only outboard valves will close D. both inboard and outboard valves will close Proposed Answer: D Explanation (Optional):

A. Incorrect: the logic arrangement of the Group 2 circuitry is such that a valid low RPV water level of 170 inches will cause an isolation.

B. Incorrect: the logic arrangement of the Group 2 circuitry is such that a valid low RPV water level of 170 inches will cause an isolation.

C. Incorrect: the logic arrangement of the Group 2 circuitry is such that a valid low RPV water level of 170 inches will cause an isolation.

D. Correct: the logic arrangement of the Group 2 circuitry is such that a valid low RPV water level of 170 inches will cause an isolation. With the conditions provided in the STEM, RPV water level is at 165 inches.

Technical Reference(s): ARP 1C05B, Revision 98 SD-959.1, Revision 13 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 93

Proposed References to be provided to applicants during examination: N Learning Objective: 50007.05.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 94

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 239002 Importance Rating Ability to predict and/or monitor changes in parameters associated with operating the RELIEF/SAFETY VALVES controls including: Reactor power. CFR: 41.5 Proposed Question: RO Question # 45 Given the following:

The plant was operating at 95% reactor power

3D Case shows the plant is operating at a 100.4% load line Subsequently, a Safety/Relief Valve received a high tailpipe temperature alarm.

The crew entered AOP 683, Abnormal Safety Relief Valve Operation

A fast power reduction to 73% was performed using recirculation flow ONLY As a result of the power reduction, which of the following is a concern?

A. SRVs should not be cycled at greater than a 90% load line B. The APRMs will be reading low as a result of the power reduction C. The power reduction may have caused the MELLLA line to be exceeded D. Core power oscillations may occur as a result of entering the Buffer Region of the Power/Flow Map Proposed Answer: C Explanation (Optional):

A. Incorrect: AOP 683 does not prescribe a flow control below other than < the MELLLA line.

B. Incorrect: The APRMs should be reading correctly.

C. Correct: AOP 683 contains a caution that the MELLLA line is expected to be exceeded, and a step requiring that power must be below the MELLLA line before a SRV is cycled.

D. Incorrect - At 75% power the reactor is well above the Buffer Region of the Power/Flow Map Technical Reference(s): SD-183.1, Revision 7 ARP 1C03A (C-5), Revision 53 AOP 683, Revision 16 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 95

Proposed References to be provided to applicants during examination: N Learning Objective: 94.51.01.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 96

Examination Outline Cross-reference: Level RO SRO Tier #

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K/A #

Importance Rating K/A: 259002A2.02: Ability to (a) predict the impacts of the following on the REACTOR WATER LEVEL CONTROL SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations:

Loss of any number of reactor feedwater flow inputs: 3..3; 3.4 Proposed Question: RO Question # 46 Given the following:

The plant is operating at 100% reactor power

Annunciator 1C05A(D-1), Reactor Vessel Hi/Lo Level Recorder Alarm, is in alarm

There are no other annunciators alarming What level do you expect the Level Recorder to indicate and how can you restore level?

A. Above 195 inches or below 170 inches. Put the A Feed REG Valve Manual/AUTO Transfer to MAN and adjust BIAS SET to restore level B. Above 195 inches or below 186 inches. Put the B Feed REG Valve Manual/AUTO Transfer to MAN and adjust BIAS SET to restore level C. Above 195 inches or below 186 inches. Put the Master Feed REG Valve AUTO/MAN Control to MAN and adjust Manual Output Adjust Knob to restore level D. Above 195 inches or below 170 inches. Put the Startup Feed REG Valve Manual/AUTO Transfer to MAN and adjust Manual Output Control to restore level Proposed Answer: C Explanation (Optional):

A. Incorrect -The level is too low; the master feed reg valve controller has to be placed in AUTO.

B. Incorrect - The level is correct; the master feed reg valve controller has to be placed in AUTO.

C. Correct D. Incorrect - The master feed reg valve controller has to be placed in AUTO.

Technical Reference(s): ARP 1C05A D 1, Rev. 1, pp. 1 3 OI 644, Rev. 27, p. 24 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 97

Proposed References to be provided to applicants during examination: N Learning Objective: 94.56.03.02 (As available)

BANK - DAEC Question Source: Bank #

NRC Exam 1994 Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 98

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 261000 A3.03 Importance Rating Ability to monitor automatic operations of the STANDBY GAS TREATMENT SYSTEM including: Valve operation. CFR: 41.7 Proposed Question: RO Question # 47 Given the following:

The Standby Gas Treatment (SBGT) system is in standby readiness condition

A complete loss of instrument air occurs The COOL DOWN dampers will fail __(1)__ and the DISCHARGE dampers will fail __(2)__ on a loss of air.

A. (1) closed (2) open B. (1) open (2) open C. (1) open (2) closed D. (1) closed (2) closed Proposed Answer: A Explanation (Optional):

A. Correct - The Instrument Air System supplies control air for the suction and discharge dampers in the system. With a loss of air the dampers will fail in a position to provide an open flow path through the filter trains. The cool down air damper fails closed and the discharge dampers fail open.

B. Incorrect - See correct answer.

C. Incorrect - See correct answer.

D. Incorrect - See correct answer.

Technical Reference(s): SD-170, Rev. 13 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 99

Proposed References to be provided to applicants during examination: N Learning Objective: 7.00.00.02 (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

The SBGT system is in standby readiness condition. What effect will a complete loss of instrument air have on the SBGT Train INTAKE, FAN INLET, and DISCHARGE valves?

These are normally _______ valves and will fail _______ on a loss of air.

A. Closed Open B. Open Closed C. Open Open D. Closed Closed Answer: C

Reference:

SD-170, Rev. 11 Answer Explanation:

Air operated ventilation dampers receive their primary air supply from the Instrument and Service Air compressors with a backup supply provided from the heating and ventilation air compressors1K-3 and 1K-4. On loss of control air, the dampers fail in such a manner as to line the SBGT System up for operation. These valves are normally in the OPEN position and OPEN upon an initiation to support operation of the system. During a complete loss of Instrument Air, the valve will fail OPEN.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 100

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A # 262001, AC Electrical Distribution, A4.05: Ability to manually operate and/or monitor in the control room: Voltage, current, power, and frequency on A.C. buses. CFR: 41.7 Proposed Question: RO Question # 48 When synchronizing 1G31, "A Standby Diesel Generator (SBDG), to the 1A3 bus, the following conditions exist:

The incoming voltage is slightly HIGHER than running voltage

The synchroscope is rotating slowly in the clockwise direction The "A" SBDG output breaker is then placed to CLOSE when the synchroscope is at the 3 o'clock position.

Which of the following describes the expected response and why?

The "A" SBDG output breaker will....

A. close and then trip open due to sensing an overspeed trip B. close and then trip open due to sensing an instantaneous overcurrent trip C. remain open due to a sync-check relay current differential D. remain open due to a sync-check relay incoming to running phase angle differential Proposed Answer: D Explanation (Optional):

A. Incorrect - An overspeed trip could occur IF the breaker closed in. The breaker will not close due to the sync-check relay action.

B. Incorrect - An instantaneous overcurrent condition due to the large phase difference IF the breaker closed. The breaker will not close due to the sync-check relay action.

C. Incorrect - The breaker remains open, however, but not due to excessive current differential.

D. Correct - The sync-check relay prevents closing in the SBDG output breaker if too large a phase difference is sensed. This protects the electrical plant from inadvertent paralleling of power sources that are not synchronized and the resulting damage that could occur.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 101

Technical Reference(s): OI 324, Rev 113 Proposed References to be provided to applicants during examination: N Learning Objective: 19.04.01.09 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 102

Examination Outline Cross-reference: Level RO SRO Tier #

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K/A #

Importance Rating K/A # 262002 UPS (AC/DC), A3.01: Transfer from preferred to alternate source, CFR 41.7 Proposed Question: RO Question # 49 The Uninterruptible AC System Transfer Switch 1Y22 will automatically transfer power from_______(1)________ to ________(2)________ on an undervoltage condition.

A. (1) 1D45 Inverter/1Y4 Regulating Transformer (2) Instrument AC Transformer 1Y2 B. (1) 1D15 Inverter/1Y1A Regulating Transformer (2) Instrument AC Transformer 1Y1 C. (1) 1D45 Inverter/1Y4 Regulating Transformer (2) Instrument AC Transformer 1Y1 D. (1) 1D15 Inverter/1Y1A Regulating Transformer (2) Instrument AC Transformer 1Y2 Proposed Answer: A Explanation (Optional):

A. Correct - System Description 357, Figure 1, pg 6, and LP 50000_357 Rev. 1, pg 14/15.

B. Incorrect - Relates to Instrument AC and not UAC.

C. Incorrect - Refers to transformer 1Y1 vice 1Y2.

D. Incorrect - Relates to Instrument AC and not UAC.

Technical Reference(s): SD-357, Rev. 7 Lesson Plan 50000_357, Rev. 1 Proposed References to be provided to applicants during examination: N Learning Objective: 21.00.00.03 (As available) 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 103

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 104

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 263000 K1.01 Importance Rating Knowledge of the physical connections and/or cause effect relationships between D.C.

ELECTRICAL DISTRIBUTION and the following: A.C. electrical distribution. CFR: 41.2 to 41.9 Proposed Question: RO Question # 50 Given the following:

The plant was operating at 100% power

4KV breaker control power has been lost to ONLY bus 1A2

AOP 302.1, Loss of 125 VDC Power, has been entered

The Control Room staff are in the process of diagnosing the failure so that the appropriate procedure section can be used Based upon the conditions above, which of the following buses has lost power?

A. 1D11, 125 VDC Division 1 Distribution Panel A B. 1D10, 125 VDC Division 1 Distribution Panel #1 C. 1D21, 125 VDC Division 2 Distribution Panel B D. 1D20, 125 VDC Division 2 Distribution Panel #2 Proposed Answer: C Explanation (Optional):

A. Incorrect: a loss of 1D11 results in a loss of 1A1 breaker control, not 1A2. (AOP-302.1 page 2)

B. Incorrect: a loss of 125 VDC DIV I (1D10) results in a loss of 1A1 and 1A3 breaker control, not 1A2. (AOP-302.1 page 2)

C. Correct: a loss of 1D21 results in a loss of breaker 1A2 control. (AOP-302.1 page 28)

D. Incorrect: a loss of 125 VDC DIV II (1D20) results in a loss of both 1A2 and 1A4 breaker control. (AOP-302.1 page 43)

Technical Reference(s): AOP-302.1, Revision 54 SD-375, Revision 8 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 105

Proposed References to be provided to applicants during examination: N Learning Objective: 13.00.00.04 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 106

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 261000 K1.07 Importance Rating Knowledge of the effect that a loss or malfunction of the EMERGENCY GENERATORS (DIESEL/JET) will have on following: A.C. electrical distribution. CFR: 41.7 Proposed Question: RO Question # 51 Given the following:

The plant was operating at 100% reactor power

A loss of offsite power has occurred

Standby Diesel Generator (SBDG) 1G31 started automatically

SBDG 1G31 running speed rose to 840 RPM

SBDG 1G31 output voltage rose to 3600 volts Which of the following describes the status of Bus 1A3?

A. Becomes energized after SBDG 1G31 trips, restarts, and closes its output breaker automatically B. Remains de-energized due to the SBDG 1G31 output breaker not closing due to low output voltage C. Becomes energized after SBDG 1G31 closes its output breaker automatically D. Remains de-energized due to the SBDG 1G31 output breaker not closing due to engine speed being too low Proposed Answer: B Explanation (Optional):

A. Incorrect: if generator output conditions are interpreted as being indicative of a fault, then generator fault protection is provided by a lockout protective feature, which acts to open the generator output breaker and also trips the engine. This feature requires manual action to reset however, and a subsequent engine restart/output breaker closure would not automatically occur in spite of the undervoltage condition still existing on bus 1A3. (SD-324 page 32).

B. Correct: the output breaker will not automatically close because output voltage is too low. Voltage must be at least 90% of rated voltage (90% of 4160V is 3744V). The voltage specified in this question is 3600V (approximately 87% of rated voltage). (SD-324 page 48).

C. Incorrect: the output breaker will not automatically close because output voltage is too low. Voltage must be at least 90% of rated voltage (90% of 4160V is 3744V). The 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 107

voltage specified in this question is 3600V (approximately 87% of rated voltage). (SD-324 page 48).

D. Incorrect: engine running speed is high enough to satisfy the required value (90% of rated speed) for output breaker closure. The rated speed of the engine is 900 RPM and 90% of this value would be 810 RPM. The running speed given in this question is 840 RPM which is approximately 93% of the rated speed. The output breaker will not automatically close because output voltage is too low. Voltage must be at least 90% of rated voltage (90% of 4160V is 3744V). The voltage specified in this question is 3600V (approximately 87% of rated voltage). (SD-324 page 48).

Technical Reference(s): SD-304, Revision 19 SD-324, Revision 15 Proposed References to be provided to applicants during examination: N Learning Objective: 19.01.01.07 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 108

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 300000 K3.02 Importance Rating Knowledge of the effect that a loss or malfunction of the INSTRUMENT AIR SYSTEM will have on the following: Systems having pneumatic valves and controls. CFR: 41.7 Proposed Question: RO Question # 52 Given the following:

The plant is operating at 100% reactor power

A total loss of instrument and service air occurred

The plant was manually scrammed

AOP 518, Failure of Instrument and Service Air was entered The CB/SBGTS Instrument Air Compressors, 1K-3 and 1K-4, will start at ___(1)___ and are NORMALLY cooled by ___(2)____.

A. (1) 90 psig (2) RBCCW B. (1) 80 psig (2) well water C. (1) 95 psig (2) well water D. (1) 80 psig (2) RBCCW Proposed Answer: B Explanation (Optional):

A. Incorrect - 1K-3 and 1K-4 air compressors will start if pressure decreases to 80 psig, to supply the system, and will stop at 88 psig. The normal cooling water to these air compressors are supplied from well water and will be cooled from Emergency Service Water (ESW) when this system is in operation.

B. Correct - 1K-3 and 1K-4 air compressors will start if pressure decreases to 80 psig, to supply the system, and will stop at 88 psig. The normal cooling water to these air compressors are supplied from well water and will be cooled from Emergency Service Water (ESW) when this system is in operation.

C. Incorrect - 1K-3 and 1K-4 air compressors will start if pressure decreases to 80 psig, to supply the system, and will stop at 88 psig.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 109

D. Incorrect - The cooling water to the 1K-3 and 1K-4 air compressors are supplied from well water and will be cooled from Emergency Service Water (ESW) when this system is in operation.

Technical Reference(s): SD-170, Rev. 13 Proposed References to be provided to applicants during examination: N Learning Objective: 7.00.00.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 110

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 400000 K4.01 Importance Rating Knowledge of CCWS design feature(s) and or interlocks which provide for the following:

Automatic start of standby pump. CFR: 41.7 Proposed Question: RO Question # 53 Given the following:

The plant was operating at 100% power

Reactor Building Closed Cooling Water (RBCCW) Pumps 1P-81A and 1P-81C were running

The supply breaker for 480 VAC Essential MCC 1B35 just tripped due to an electrical fault Based on these conditions, when RBCCW pressure LOWERS to A. 55 psig RBCCW pumps 1P-81A and 1P-81B will be running.

B. 55 psig RBCCW pumps 1P-81B and 1P-81C will be running.

C. 35 psig RBCCW pumps 1P-81A and 1P-81B will be running.

D. 35 psig RBCCW pumps 1P-81B and 1P-81C will be running.

Proposed Answer: D Explanation (Optional):

A. Incorrect: the low pressure auto-start of the standby pump occurs at 35 psig, while 55 psig corresponds to the auto start pressure of a General Service Water Pump. Also, the electrical power supply for RBCCW Pump 1P-81A is 480 VAC Essential MCC 1B35, whereas RBCCW Pumps 1P-81B and 1P-81C are supplied from 480 VAC Essential MCC 1B43 (SD-414 page 7).

B. Incorrect: the low pressure auto-start of the standby pump occurs at 35 psig, while 55 psig corresponds to the auto start pressure of a General Service Water Pump.

C. Incorrect: the electrical power supply for RBCCW Pump 1P-81A is 480 VAC Essential MCC 1B35, whereas RBCCW Pumps 1P-81B and 1P-81C are supplied from 480 VAC Essential MCC 1B43 (SD-414 page 7).

D. Correct.

Technical Reference(s): ARP 1C06B, Revision 56 SD-414, Revision 9 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 111

Proposed References to be provided to applicants during examination: N Learning Objective: 29.00.00.04 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 112

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 201001 K2.02 Importance Rating Knowledge of electrical power supplies to the following: Scram valve solenoids. CFR: 41.7 Proposed Question: RO Question # 54 Given the following:

The plant is operating at 80% power

Reactor Protection System Buses are being supplied by their normal sources

An electrical fault results in a loss of power to Essential 480 VAC Motor Control Center 1B42

Subsequently, Reactor Vessel High Pressure Trip Channel A1 experiences a spurious trip Based on the conditions above, what will be the status of ALL scram pilot valves SV-1855 and SV-1856?

A. Both energized B. Only SV-1855 valves will be de-energized C. Only SV-1856 valves will be de-energized D. Both de-energized Proposed Answer: D Explanation (Optional):

A. Incorrect: the loss of MCC 1B42 resulted in a loss of RPS MG Set B. This would cause a loss of power to RPS train B, which results in a half scram condition. Scram pilot valve solenoids SV-1855 will lose power (SD-255 pages 24-25, and SD-358 pages 10-29). Also, the trip of a single Reactor Vessel High Pressure Trip Channel results in a half scram (ARP 1C05B C-4). Since this occurred on the A1 channel, this would result in a RPS train A half scram and the SV-1856 scram pilot valves would de-energize (SD-255 pages 24-25). Based on this combination of events, both the SV-1856 and SV-1855 scram pilot valves have been de-energized, resulting in a full scram condition.

B. Incorrect: the trip of a single Reactor Vessel High Pressure Trip Channel results in a half scram (ARP 1C05B C-4). Since this occurred on the A1 channel, this would result in a RPS train A half scram and the SV-1856 scram pilot valves would de-energize (SD-255 pages 24-25).

C. Incorrect: the loss of MCC 1B42 resulted in a loss of RPS MG Set B. This would cause a loss of power to RPS train B, which results in a half scram condition. Scram 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 113

pilot valve solenoids SV-1855 will lose power (SD-255 pages 24-25, and SD-358 pages 10-29).

D. Correct: the loss of MCC 1B42 resulted in a loss of RPS MG Set B. This would cause a loss of power to RPS train B, which results in a half scram condition. Scram pilot valve solenoids SV-1855 will lose power (SD-255 pages 24-25, and SD-358 pages 10-29). Also, the trip of a single Reactor Vessel High Pressure Trip Channel results in a half scram (ARP 1C05B C-4). Since this occurred on the A1 channel, this would result in a RPS train A half scram and the SV-1856 scram pilot valves would de-energize (SD-255 pages 24-25). Based on this combination of events, both the SV-1856 and SV-1855 scram pilot valves have been de-energized, resulting in a full scram condition.

Technical Reference(s): ARP 1C05B, Revision 98 SD-255, Revision 9 SD-358, Revision 9 Proposed References to be provided to applicants during examination: N Learning Objective: 22.00.00.04 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 114

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 202001 2.2.25 Importance Rating Knowledge of the bases in Technical Specifications for limiting conditions for operations and safety limits. CFR: 41.5 / 41.7. Associated system: Recirculation Proposed Question: RO Question # 55 The recirculation loop speed mismatch limits of TS 3.4.1, Recirculation Loops Operating, are based upon preventing which of the following during a subsequent Loss of Coolant Accident scenario?

A. Ability to re-flood the core to 2/3 core height post LOCA B. Assumed blowdown flow being invalidated C. LPCI Loop Select Logic selecting the wrong loop D. Excessive flow coastdown characteristics Proposed Answer: C Explanation (Optional):

A. Incorrect: this adverse effect is associated with jet pump operability. The ITS 3.4.2 Basis states the following: The capability of reflooding the core to two-thirds core height is dependent upon the structural integrity of the jet pumps. If the structural system, including the beam holding a jet pump in place, fails, jet pump displacement and performance degradation could occur, resulting in an increased flow area through the jet pump and a lower core flooding elevation. This could adversely affect the water level in the core during the reflood phase of a LOCA as well as the assumed blowdown flow during a LOCA.

B. Incorrect: this adverse effect is associated with jet pump operability. The ITS 3.4.2 Basis states the following: The capability of reflooding the core to two-thirds core height is dependent upon the structural integrity of the jet pumps. If the structural system, including the beam holding a jet pump in place, fails, jet pump displacement and performance degradation could occur, resulting in an increased flow area through the jet pump and a lower core flooding elevation. This could adversely affect the water level in the core during the reflood phase of a LOCA as well as the assumed blowdown flow during a LOCA.

C. Correct: the basis of ITS 3.4.1 states the following: Since recirculation loop flow is controlled by varying recirculation pump speed, a limit on the speed mismatch between operating recirculation pumps has been imposed. For some limited low probability accidents (e.g., intermediate break size LOCAs) with the recirculation loop operating with large speed differences, it is possible for the LPCI Loop Select Logic to select the wrong loop for injection. For these limited conditions the Core Spray itself is adequate 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 115

to prevent fuel temperatures from exceeding allowable limits. However, to limit the probability even further, operating procedures have been put into place limiting the allowable mismatch in speed between the recirculation pumps. Analyses indicate that above 69.4% RTP the Loop Select Logic could be expected to function at a speed differential up to 14% of their average speed. Below 69.4% RTP the Loop Select Logic would be expected to function at a speed differential up to 20% of their average speed.

The recirculation loop speed mismatch limits imposed to prevent the LPCI Loop Select Logic from selecting the wrong loop for injection bound the recirculation flow mismatch limits for LOCA analyses.

D. Incorrect: the ITS 3.4.1 Basis states the following: The recirculation loop speed mismatch limits imposed to prevent the LPCI Loop Select Logic from selecting the wrong loop for injection bound the recirculation flow mismatch limits for LOCA analyses.

If the reactor is operating on one recirculation pump, the Loop Select Logic trips that pump before making the loop selection. The recirculation system is also assumed to have sufficient flow coastdown characteristics to maintain fuel thermal margins during abnormal operational transients (Ref. 2), which are analyzed in Chapter 15 of the UFSAR. This answer is incorrect because of the key wording excessive flow coastdown. The concern per the basis would be insufficient flow coastdown, not excessive flow coastdown.

Technical Reference(s): DAEC ITS Bases 3.4.1 DAEC ITS Bases 3.4.2 Proposed References to be provided to applicants during examination: N Learning Objective: 12.00.00.06 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 116

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 204000 K4.03 Importance Rating Knowledge of REACTOR WATER CLEANUP SYSTEM design feature(s) and/or interlocks which provide for the following: Over temperature protection for system components. CFR:

41.7 Proposed Question: RO Question # 56 Given the following:

The plant was operating at 90% power

Reactor Water Cleanup Pump 1P-205A is running

Subsequently, Reactor Building Closed Cooling Water (RBCCW) flow was lost to the Reactor Water Cleanup (RWCU) system

1C04B (C-9), RWCU Filter/DEMIN Inlet Water HI Temp, is now LIT

RWCU Filter/Demin Inlet temperature is 130°F and rising

NO operator action has yet been taken Which of the following describes the CURRENT status of the RWCU system?

A. MO 2700, INBD Cleanup SUCT ISOL, is OPEN MO 2701, OUTBD Cleanup SUCT ISOL, is CLOSED B. MO 2700, INBD Cleanup SUCT ISOL, is CLOSED MO 2701, OUTBD Cleanup SUCT ISOL, is CLOSED C. MO 2700, INBD Cleanup SUCT ISOL, is OPEN MO 2701, OUTBD Cleanup SUCT ISOL, is OPEN D. MO 2700, INBD Cleanup SUCT ISOL, is CLOSED MO 2701, OUTBD Cleanup SUCT ISOL, is OPEN Proposed Answer: C Explanation (Optional):

A. Incorrect: MO 2701 closes at a Filter/demineralizer inlet High Temperature at 140ºF; this feature affects outboard isolation valves, and not inboard isolation valves (SD-261 page 11 and ARP 1C04B D-9). This temperature would also correspond to the receipt of 1C04B (D-9), RWCU FILTER/DEMIN INLET WATER HI-HI TEMP (ARP 1C04B C-9).

B. Incorrect: MO 2701 closes at a Filter/demineralizer inlet High Temperature at 140ºF; this feature affects outboard isolation valves, and not inboard isolation valves (SD-261 page 11 and ARP 1C04B D-9). This temperature would also correspond to the receipt of 1C04B (D-9), RWCU FILTER/DEMIN INLET WATER HI-HI TEMP (ARP 1C04B C-9).

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 117

C. Correct: MO 2701 closes at a Filter/demineralizer inlet High Temperature at 140ºF (SD-261 page 11 and ARP 1C04B D-9).

D. Incorrect: MO 2701 closes at a Filter/demineralizer inlet High Temperature at 140ºF; this feature affects outboard isolation valves, and not inboard isolation valves (SD-261 page 11 and ARP 1C04B D-9). This temperature would also correspond to the receipt of 1C04B (D-9), RWCU FILTER/DEMIN INLET WATER HI-HI TEMP (ARP 1C04B C-9).

Technical Reference(s): ARP 1C04B SD-261 Proposed References to be provided to applicants during examination: N Learning Objective: 11.00.00.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 118

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 214000 Importance Rating Knowledge of the effect that a loss or malfunction of the ROD POSITION INFORMATION SYSTEM will have on the following: RMCS Proposed Question: RO Question # 57 A reactor shutdown is in progress with the plant at 50% power when an RMCS malfunction forces you to use EMERG IN for rod insertion.

Which of the following would prevent use of EMERG IN?

A. A bypassed Rod Worth Minimizer B. ROD OUT BLOCK annunciator 1C05B(A-6) in alarm C. RBM UPSCALE OR INOP annunciator 1C05B(B-6) in alarm D. No position indication for the currently selected control rod due to a failed reed switch Proposed Answer: D Explanation (Optional):

A. Incorrect: A bypassed Rod Worth Minimizer will not prevent use of EMERG IN.

B. Incorrect: A ROD OUT BLOCK will not prevent use of EMERG IN.

C. Incorrect: A RBM UPSCALE OR INOP will not prevent use of EMERG IN.

D. Correct: A reed switch failure will result in all control rods receiving an insert and withdrawal block. This block is generated from the RWM and is not bypassed by the EMERGENCY IN switch operation. The RWM must be BYPASSED for rod insertion to continue.

Technical Reference(s): SD-878.8, Rev. 8 Proposed References to be provided to applicants during examination: N Learning Objective: 84.00.00.02 (As available) 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 119

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 120

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A # 215002 RBM A4.03: Ability to manually operate and/or monitor in the control room: Trip bypasses: BWR-3,4,5. CFR: 41.7 Proposed Question: RO Question # 58 The plant is operating at 100% power under the following conditions:

Repairs on "A" Rod Block Monitor (RBM) were completed

RBM "A" was removed from BYPASS to accomplish Post Maintenance Testing

The ROD OUT PERMISSIVE light extinguished and illuminated again two seconds later

Annunciator 1C05B(A-6), Rod Out Block, did NOT alarm Which statement below correctly describes the response to the given conditions?

This response was.

A. NOT normal because the A RBM should NOT null until a new control rod is selected B. normal because "A" RBM generated a rod out inhibit during the null sequence C. NOT normal only because the annunciator should have alarmed when the ROD OUT PERMISSIVE light was extinguished D. normal because the rod out blocks are bypassed for two seconds to allow the reference APRM gain adjustment during the null sequence Proposed Answer: B Explanation (Optional):

A. Incorrect -.Taking the RBM out of BYPASS will initiates a null sequence.

B. Correct - Taking a RBM out of BYPASS initiates a null sequence. RBM trip functions are bypassed during the nulling sequence so no alarm is generated.

C. Incorrect - The RBM trip functions are bypassed during the nulling sequence so no alarm is generated.

D. Incorrect - There is no rod block bypass, the RBM trip functions are bypassed during the nulling sequence so no alarm is generated.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 121

Technical Reference(s): SD-878.5, Rev. 10, Page 16 Proposed References to be provided to applicants during examination: N Learning Objective: 82.00.00.02 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 2011 DAEC NRC Question History: Last NRC Exam:

Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 122

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A # 216000 Nuclear Boiler Inst. K6.03: Knowledge of the effect that a loss or malfunction of the following will have on the NUCLEAR BOILER INSTRUMENTATION: Temperature compensation. CFR: 41.7 Proposed Question: RO Question # 59 Which of the following instruments are TEMPERATURE compensated AND calibrated HOT?

A. Wide Range Yarway Level Instruments B. Narrow Range GEMAC Level Instruments C. Wide Range GEMAC Floodup Instruments D. Fuel Zone Instruments Proposed Answer: A Explanation (Optional):

A. Correct.

B. Incorrect. Narrow Range GEMAC level transmitters are NOT compensated but are calibrated HOT.

C. Incorrect. Wide Range GEMAC Floodup Instruments are NOT compensated and are also calibrated COLD.

D. Incorrect. Electronic pressure compensation and COLD calibration are used for the Fuel Zone Instruments.

50007_88-0_Part 1_lp page 11-Technical Reference(s): SD-880 rev 13 page 27 16 Proposed References to be provided to applicants during examination: N Learning Objective: 88.00.00.02 (As available)

Question Source: Bank #

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 123

Modified Bank # X (Note changes or attach parent)

New ILT NRC 2001 Question History: Last NRC Exam:

Examination DAEC Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 124

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 233000 A1.03 Importance Rating Ability to predict and/or monitor changes in parameters associated with operating the FUEL POOL COOLING AND CLEAN-UP controls including: Pool temperature. CFR: 41.5 Proposed Question: RO Question # 60 Given the following:

The plant was operating at 100%

The running Fuel Pool Cooling Pump, 1P-214A, tripped due to a seized bearing

While subsequently attempting to start Fuel Pool Cooling Pump 1P-214B using hand switch HS 3410B in accordance with OI-435, Fuel Pool Cooling System, an electrical fault in the control circuit prevented pump start

Fuel Pool Temperature was initially 70 degrees

Fuel Pool heatup rate has been calculated to be 2.1°F/hr How long will Fuel Pool Temperature take to rise to the maximum design limit?

A. 15.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months B. 24.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months C. 28.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months D. 38.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months Proposed Answer: D Explanation (Optional):

A. Incorrect: this answer incorrectly assumes that the minimum operating limit is 68°F (this limit applies when fuel pool gates are removed; the fact that these gates are installed is provided indirectly by the plant condition of 100% in the stem) (SD-435 page 10). It also incorrectly assumes that the maximum operating limit is 130°F (which is the heat exchanger outlet temperature limit to prevent resin damage) (SD-435 page 10). These assumptions yield the following calculation: [130°F - (68°F + 30°F)] / 2.1°F/hour = 15.2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. Incorrect: this answer incorrectly assumes that the minimum operating limit is 68°F (this limit applies when fuel pool gates are removed; the fact that these gates are installed is provided indirectly by the plant condition of 100% in the stem) (SD-435 page 10). This assumption yields the following calculation: [150°F - (68°F + 30°F)] / 2.1°F/hour = 24.8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

C. Incorrect: this answer incorrectly assumes that the maximum operating limit is 130°F (the heat exchanger outlet temperature limit to prevent resin damage) (SD-435 page 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 125

10). This assumption yields the following calculation: [130°F - (40°F + 30°F)] /

2.1°F/hour = 28.6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

D. Correct: the minimum fuel pool operating limit is 40°F when fuel pool gates are installed (SD-435 page 10). The maximum fuel pool operating limit is 150°F (SD-435 page 10).

These assumptions yield the following calculation: [150°F - (40°F + 30°F)] / 2.1°F/hour

= 38.1 hours1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months .

Technical Reference(s): AOP-435, Revision 10 ARP 1C04B, Revision 79 OI-435, Revision 65 SD-435, Revision 8 Proposed References to be provided to applicants during examination: N Learning Objective: 31.00.00.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 126

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 234000 K1.01 Importance Rating Knowledge of the physical connections and/or cause effect relationships between FUEL HANDLING EQUIPMENT and the following: Fuel. CFR: 41.2 to 41.9 Proposed Question: RO Question # 61 Given the following:

Refueling operations are in progress

The refueling platform operator is in the process of removing a fuel assembly from the reactor

The grapple and fuel assembly have just been raised

Both the GRAPPLE ENGAGED and GRAPPLE NORMAL UP lights are LIT

Suddenly a leak develops in the air supply to the grapple When the air supply pressure to the grapple lowers to 90 psig, the status of the GRAPPLE ENGAGED light will be ___(1)___ and the position of the grapple will be ___(2)____.

A. (1) NOT lit (2) open B. (1) NOT lit (2) closed C. (1) LIT (2) open D. (1) LIT (2) closed Proposed Answer: D Explanation (Optional):

A. Incorrect: the GRAPPLE ENGAGED light will illuminate based upon the grapple state as sensed by limit switches. If it is not lit, the hooks of the fuel grapple are not fully closed.

When the grapple senses less than 100 psi in its air supply line, the grapple automatically closes (SD-281 pages 12-13). Thus the grapple has not dropped the fuel assembly. The light will continue to indicate the true state of the grapple, which remains engaged.

B. Incorrect: the GRAPPLE ENGAGED light will illuminate based upon the grapple state as sensed by limit switches. If it is not lit, the hooks of the fuel grapple are not fully closed (SD-281 pages 12-13). The light will continue to indicate the true state of the grapple, 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 127

which remains engaged.

C. Incorrect: the GRAPPLE ENGAGED light will illuminate based upon the grapple state as sensed by limit switches. If it is not lit, the hooks of the fuel grapple are not fully closed.

When the grapple senses less than 100 psi in its air supply line, the grapple automatically closes (SD-281 pages 12-13). Thus the grapple has not dropped the fuel assembly. The light will continue to indicate the true state of the grapple, which remains engaged.

D. Correct.

Technical Reference(s): SD-281, Revision 7 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 128

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 268000 G2.4.21 Importance Rating 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.

Proposed Question: RO Question # 62 The plant is operating in MODE 1 at 100% power with the following conditions:

Annunciator 1C147(A-2), Reactor BLDG South East Area Floor Drain Level High, is in alarm

Annunciator 1C14A(B-4), Area Water Levels Above MAX Normal, is in alarm

An operator reports from 1C21 that SE Corner Room (SECR) level is slightly greater than 2 inches, at Max Normal, and rising very slowly

There are SECR mezzanine reports of water on the floor and they are trying to locate the leak Which of the following procedures:

(1) Shall be reported to the CRS for possible entry into?

AND (2) What are the required actions?

A. (1) EOP 2, Primary Containment Control (2) Scram the reactor and emergency depressurize B. (1) EOP 3, Secondary Containment Control (2) Have the Plant Chemist sample the water prior to draining it to the Reactor Building Floor Drain Sump C. (1) EOP 2, Primary Containment Control (2) Have the Radwaste Operator pump down the Reactor Building Floor Drain Sump D. (1) EOP 3, Secondary Containment Control (2) Have the Radwaste Operator open the affected valve to drain the area, and operate sump pumps as necessary Proposed Answer: D Explanation (Optional):

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 129

A. Incorrect - The greater than max normal water level is an entry into EOP 3, not EOP 2.

B. Incorrect - There is no requirement to sample the water and time should not be spent in the EOP sampling the discharge of water from this area is required.

C. Incorrect - The greater than max normal water level is an entry into EOP 3, not EOP 2.

D. Correct - SE Corner Room level is slightly greater than 2 inches is above the Max Normal Operating Limit for the SE corner Room which requires an entry into EOP-3.

The EOP requires operating available sump pumps to restore and maintain water level below the Max Normal Operating Limit Technical Reference(s): Bases-EOP 3, Rev.

Proposed References to be provided to applicants during examination: N Learning Objective: 95.68.01.01 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2011 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

The plant is operating in MODE 1 at 100% power with the following conditions:

Annunciator A-2 REACTOR BLDG SOUTH EAST AREA FLOOR DRAIN LEVEL

HIGH alarms at panel 1C147, RB Floor Drain System Control

Annunciator B-4 AREA WATER LEVELS ABOVE MAX NORMAL alarms at panel 1C14A, EOP Annunciators

An operator reports from 1C21 that SE Corner Room level is slightly greater than 2 inches and rising very slowly.

SANSOE reports from the SECR mezzanine that there is water on the floor and he will try to locate the leak Which one of the following procedures:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 130

(1) Shall be reported to the CRS as a possible entry, and (2) What are the required actions A. (1) EOP 1, RPV CONTROL (2) Scram the reactor and control level, pressure, reactor power.

B. (1) EOP 3, SECONDARY CONTAINMENT CONTROL (2) Contact the Plant Chemist and have him sample the water prior to draining it to the Reactor Building Floor Drain Sump.

C. (1) EOP 1, RPV CONTROL (2) Contact the Radwaste Operator and have him pump down the Reactor Building Floor Drain Sump.

D. (1) EOP 3, SECONDARY CONTAINMENT CONTROL (2) Have the Radwaste Operator open the affected valve to drain the area, and operate sump pumps as necessary.

Proposed Answer: D 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 131

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A # 271000 Offgas, A3.05: Ability to monitor automatic operations of the OFFGAS SYSTEM including: System indicating lights and alarms. CFR: 41.7 Proposed Question: RO Question # 63 The plant has been scrammed with indications of fuel damage.

The following annunciators are in alarm:

1C05B(D-2), Main Steam Line HI RAD

1C05B(C-2), Main Steam Line HI HI RAD / INOP trip

1C03A(B-4), Offgas Vent Pipe RM-4116A/B HI RAD

1C03A(A-4), Offgas Vent Pipe RM-4116A/B HI-HI RAD Based on the alarms and conditions above, which of the following isolations are NOT expected to occur AUTOMATICALLY?

A. Recirc Sample Valves B. PCIS Group III C. Main Steam Line Drains D. Main Steam Line Isolation Valves (MSIVs)

Proposed Answer: D Explanation (Optional):

A. Incorrect: Recirc Sample Control Valves will automatically isolate.

B. Incorrect. A Group III isolation is expected.

C. Incorrect. Main Steam Line Drains will automatically isolate.

D. Correct. The MSIVs must be manually isolated.

Technical Reference(s): AOP 672, Rev.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 132

Proposed References to be provided to applicants during examination: N Learning Objective: 50007.05.02 (As available)

Question Source: Bank #

2009 DAEC NRC Modified Bank # (Note changes or attach parent)

Exam New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 133

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A # 272000, Radiation Monitoring, K2.05: Knowledge of the operational implications of the following concepts as they apply to RADIATION MONITORING SYSTEM: Reactor building ventilation monitors: Plant-Specific. CFR: 41.7 Proposed Question: RO Question # 64 The plant is operating at power with all LCOs met when annunciator 1C23A(F-3), Reactor BLDG Vent Shaft RAD Monitor RIM-7606A HI/Trouble, IS in alarm.

Which one of the following describes:

(1) a potential cause of the alarm?

AND (2) a resulting automatic action that occurs due to the alarm, if any?

A. (1) Loss of its Instrument AC power supply (2) NO actions B. (1) Loss of its Instrument AC power supply (2) Inboard Group 3 isolation C. (1) Loss of its 125 VDC power supply (2) NO actions D. (1) Loss of its 125 VDC power supply (2) Inboard Group 3 isolation Proposed Answer: B Explanation (Optional):

A. Incorrect: An Inboard Group 3 isolation occurs.

B. Correct. Loss of 120VAC Instrument power causes the alarm and an Inboard Group 3 isolation occurs.

C. Incorrect. Power is supplied via 120 VAC.

D. Incorrect. Power is supplied via 120 VAC and an Inboard Group 3 isolation occurs Technical Reference(s): AOP 317, LOSS OF 120 VAC 50007_879-1_lp rev 0, pages 36, 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 134

INSTRUMENT CONTROL 37 & 58 POWER PANEL 1Y11 rev 96, p.

12 SD 879.1, PROCESS RADIATION MONITORING SYSTEM, rev 9, pages 50-51 ARP 1C23A rev 18 (F-3) Sections 1 and 3, page 61 Proposed References to be provided to applicants during examination: N Learning Objective: 85.00.00.04 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 2009 DAEC Audit Question History: Last NRC Exam:

Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 135

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 290003 K3.04 Importance Rating Knowledge of the effect that a loss or malfunction of the CONTROL ROOM HVAC will have on following: Control room pressure. CFR: 41.7 Proposed Question: RO Question # 65 Given the following:

The plant is operating at 100% reactor power

A radiological event resulted in an automatic Control Building Isolation

Five (5) minutes later, Battery Exhaust Fans 1V-EF-30A AND 1V-EF-30B are observed to remain running Based on the above conditions, the Control Building Isolation will .

A. be able to maintain the required positive Control Room pressure B. be able to maintain the required negative Control Room pressure C. NOT be able to maintain the required positive Control Room pressure D. NOT be able to maintain the required negative Battery Room pressure Proposed Answer: C Explanation (Optional):

A. Incorrect: the Control Building Isolation is designed to maintain a positive pressure in the Control Room. In order to maintain a positive pressure, only one battery exhaust fan can be running. To achieve this, the Control Building Isolation will automatically shift the three battery exhaust fans to a configuration that only leaves one running.

Since that shift failed to occur automatically in this scenario, a positive Control Room pressure will not be maintained (OI-730 page 4, SD-730 page 26).

B. Incorrect: the stem presents conditions in which a Control Building Isolation has automatically occurred. This isolation is designed to maintain a positive pressure in the Control Room (versus a negative pressure) (OI-730 page 4, SD-730 page 34).

C. Correct.

D. Incorrect: the Control Building Isolation is designed to maintain a positive pressure in the Control Room. In order to maintain a positive pressure, only one battery exhaust fan can be running. To achieve this, the Control Building Isolation will automatically shift the three battery exhaust fans to a configuration that only leaves one running.

Since that shift failed to occur automatically in this scenario, a positive Control Room pressure will not be maintained (OI-730 page 4, SD-730 page 26). Exhaust flow from 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 136

the Battery Room would continue however since battery fans continue to operate.

Technical Reference(s): ARP1C26A, Revision 50 ARP 1C26B, Revision 50 ARP 1C07A, Rev. 51 OI-730, Revision 117 SD-730, Revision 12 Proposed References to be provided to applicants during examination: N Learning Objective: 65.01.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 137

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 2.1.32 Importance Rating Ability to explain and apply system limits and precautions. CFR: 41.10 Proposed Question: RO Question # 66 Given the following for Standby Diesel Generator (SBDG) 1G-31:

The SBDG is being run for testing

The electrical output is 250 KW

The engine crankcase pressure is negative (-)0.50 inches of water

The turbocharger inlet temperature is 800°F Which of the following represents a concern associated with extended operation of SBDG 1G-31 under these conditions?

A. Crankcase explosion due to explosive gas accumulation B. Exhaust system fire due to combustion product buildup C. Excessive turbocharger wear due to overheating D. Fuel injector failure due to incomplete combustion Proposed Answer: B Explanation (Optional):

A. Incorrect: high crankcase pressure indicates the possible existence of an explosive gas mixture, with the possibility of a crankcase explosion (as discussed in the Precautions of OI-324). The crankcase pressure switches actuate at 0.5" water pressure. While the engine crankcase is normally maintained at a slightly negative pressure during operation, the value provided in the stem does not yet present an explosion hazard (OI-324 page 5, SD-324 page 11).

B. Correct: the KW loading value provided (250 KW) represents 20% of the rated load (3250 KW) of a SBDG (SD-324 page 7). OI-324 contains the following Precaution:

Avoid prolonged periods of operation at less than 25% load to avoid buildup of incomplete combustion products in the exhaust lines (engine souping), with the possibility of fire upon return to full load (OI-324 page 5).

C. Incorrect: OI-324 contains Precautions that Turbocharger inlet temperature should not exceed 1200°F, and that Diesel engine exhaust temperature shall not exceed 1100°F (OI-324 page 5). The turbocharger inlet temperature provided (800°F) is below the correct limit of 1200°F.

D. Incorrect: OI-324 contains a precaution against prolonged periods of operation at less 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 138

than 25% load to avoid buildup of incomplete combustion products (OI-324 page 5).

The component of concern is the exhaust lines however, and not the fuel injectors.

Technical Reference(s): OI-324, Revision 113 SD-324, Revision 15 Proposed References to be provided to applicants during examination: N Learning Objective: 19.00.00.03 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 139

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A # G2.1.39: Knowledge of conservative decision making practices. CFR: 41.10 Proposed Question: RO Question # 67 Given the following:

The plant was operating normally at 100% reactor power

The operator observes from SPDS data that scram criteria is met Based on the information above, what would be the proper response?

A. Scram the reactor immediately B. Inform the CRS, and then scram the reactor C. Validate SPDS data with permanent plant instrumentation, inform and with the CRSs permission, then scram the reactor D. Validate SPDS data with permanent plant instrumentation, and then scram the reactor Proposed Answer: D Explanation (Optional):

A. Incorrect: Per ACP 1410.1, no emergency action will be taken based on the SPDS data alone B. Incorrect: Per ACP 1410.1, no emergency action will be taken based on the SPDS data alone AND CRS consultation is not required C. Incorrect: CRS consultation is not required - Any on-shift RO or SRO has the authority to reduce power or shutdown the reactor when it is determined that the safety of the reactor is in jeopardy D. Correct: Any on-shift RO or SRO has the authority to reduce power or shutdown the reactor when it is determined that the safety of the reactor is in jeopardy ACP 1410.1, Rev. 100, pages 21, Technical Reference(s):

24 and 25.

Proposed References to be provided to applicants during examination: N 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 140

Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 141

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 2.2.17 Importance Rating Knowledge of the process for managing maintenance activities during power operations, such as risk assessments, work prioritization, and coordination with the transmission system operator. CFR: 41.10 Proposed Question: RO Question # 68 Given the following:

The plant is in IPOI 2, Startup, and is placing the main generator on the grid What notification is required when placing the Generator Alterrex Supplementary Control in service and when is it required to be completed by?

A. Notify ITC Midwest within 30 minutes B. Notify ITC Midwest within 60 minutes C. Notify ITC Midwest immediately D. No notifications are required Proposed Answer: A Explanation (Optional):

A. Correct - OI 698 Step 19 states that notification to ITC Midwest and Real Time Desk that the Generator Alterrex Supplementary control (Power System Stabilizer) is ON.

Additionally, a NOTE is contained in OI 698 that states that this notification is required within 30 minutes.

B. Incorrect - OI 698 contains a NOTE that states that this notification is required within 30 minutes. 60 minutes is too long.

C. Incorrect - OI 698 contains a NOTE that states that this notification is required within 30 minutes. Immediately is not required.

D. Incorrect - OI 698 Step 19 states that notification to ITC Midwest and Real Time Desk that the Generator Alterrex Supplementary control (Power System Stabilizer) is ON.

Technical Reference(s): ODI-032, Revision 5 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 142

Proposed References to be provided to applicants during examination: N Learning Objective: 57.02.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 143

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A # G2.2.2: Ability to manipulate the console controls as required to operate the facility between shutdown and designated power levels. CFR: 41.6 / 41.7 Proposed Question: RO Question # 69 A reactor startup is in progress.

Conditions at the beginning of the startup and currently are listed below:

Beginning of Startup Currently

(@100% Rod Density) (@80% Rod Density)

Channel Counts Counts SRM A 9 85 SRM B 11 100 SRM C 8 90 SRM D 10 95 The reactor is NOT critical and there is one rod left to pull to complete the current group.

In order to pull this control rod to continue the startup, what must be done per IPOI-2, Startup?

A. Continue using continuous withdrawal until the current group is complete, then use single notch withdrawal until 75% rod density is achieved B. Change to single notch withdrawal immediately and continue with single notch withdrawal until 75% rod density is achieved C. Change to single notch withdrawal until the current group is complete, then resume continuous rod withdrawal for the next group until the reactor is critical D. Continue using continuous withdrawal until all SRM count rates have increased by a factor of 10, then switch to single notch withdrawal until the reactor is critical Proposed Answer: B Explanation (Optional):

A. Incorrect - Must switch to single notch withdrawal B. Correct - Per IPOI step 27(b) - If any SRM count rate has increased by a factor of 10 prior to reaching 75% rod density, then conduct single rod notch withdrawal until the 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 144

75% rod density is achieved (all RWM Group 2 rods full out). In this case SRM "C" has increased by a factor of 10.

C. Incorrect - Must switch to single notch withdrawal and remain in single notch withdrawal until the 75% rod density is achieved.

D. Incorrect - Must switch to single notch withdrawal Technical Reference(s): IPOI-2, Rev.140 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2001 NRC Exam Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 145

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 2.2.22 Importance Rating Knowledge of limiting conditions for operations and safety limits. CFR: 41.5 Proposed Question: RO Question # 70 Given the following:

The plant was operating at 100% power when a severe transient occurred

During the transient 'B' Recirculation Pump tripped

'A' Recirculation Pump remains in operation Based upon the conditions above, which of the following describes the Minimum Critical Power Ratio Safety Limit that currently applies?

A. 1.10 while operating at <10% rated core flow B. 1.10 while operating at 10% rated core flow C. 1.12 while operating at <10% rated core flow D. 1.12 while operating at 10% rated core flow Proposed Answer: D Explanation (Optional):

A. Incorrect: MCPR shall be 1.12 for single recirculation loop operation with core flow 10% rated core flow. 1.10 is the two recirculation loop safety limit (TS Safety Limit 2.1.1).

B. Incorrect: MCPR shall be 1.12 for single recirculation loop operation. 1.10 is the two recirculation loop safety limit (TS Safety Limit 2.1.1).

C. Incorrect: MCPR shall be 1.12 for single recirculation loop operation with core flow 10% rated core flow (TS Safety Limit 2.1.1).

D. Correct.

Technical Reference(s): TS 2.1.1, Amendment 243 TS Bases 2.1.1, Amendment 223 UFSAR, Revision 18 Proposed References to be provided to applicants during examination: N 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 146

Learning Objective: 1.03.03.04 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 147

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 2.3.4 Importance Rating Knowledge of radiation exposure limits under normal or emergency conditions. CFR: 41.12 Proposed Question: RO Question # 71 Given the following:

A major plant event has occurred resulting in highly elevated radiation levels in the power block and the declaration of a General Emergency

There is an injured man pinned in the reactor building and personnel are needed to rescue the individual person Assuming that the rescue personnel are NOT volunteers, what is the maximum radiation exposure that may be authorized for the personnel performing this rescue activity?

A. >25 REM B. 25 REM C. >10 REM but <25 REM D. 10 REM Proposed Answer: B Explanation (Optional):

A. Incorrect: >25 Rem may be authorized for lifesaving or protection of large populations, but only on a voluntary basis to persons fully aware of the risks involved (Form OSC-13).

B. Correct: 25 Rem is the maximum that may be authorized for life-saving or protection of large populations (Form OSC-13).

C. Incorrect: 10 Rem is the maximum that may be authorized for the protection of valuable equipment. 25 Rem is the maximum that may be authorized for life-saving or protection of large populations (Form OSC-13).

D. Incorrect: 10 Rem is the maximum that may be authorized for the protection of valuable equipment (Form OSC-13).

Technical Reference(s): Form OSC-13, Rev. 0 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 148

Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 149

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 2.3.13 Importance Rating Knowledge of radiological safety procedures pertaining to licensed operator duties, such as response to radiation monitor alarms, containment entry requirements, fuel handling responsibilities, access to locked high-radiation areas, aligning filters, etc. CFR: 41.12 Proposed Question: RO Question # 72 Given the following:

The plant is shutdown

At the request of the Operations Department, Health Physics (HP) has established a temporary Locked High Radiation Area (LHRA)

Operations Department work is expected to be conducted in this area for the next week

An operator was briefed on the job and on the associated RWP is now seeking entry into the temporary LHRA Based upon the conditions above, ___(1)____ can grant the operator permission to enter the Temporary LHRA, and ___(2)____ can issue the key.

A. (1) EITHER HP Department or the OSM/CRS (2) ONLY the HP Department B. (1) ONLY the HP Department (2) ONLY the HP Department C. (1) ONLY the HP Department (2) EITHER HP Department or the OSM/CRS D. (1) EITHER HP Department or the OSM/CRS (2) EITHER HP Department or the OSM/CRS Proposed Answer: B Explanation (Optional):

A. Incorrect: 1st part wrong, 2nd part correct. Under emergency conditions, the OSM/CRS can grant.

B. Correct.

C. Incorrect: 1st part correct, 2nd part wrong. The OSM/CRS maintains a set of LHRA Master Keys.

D. Incorrect: 1st part wrong, 2nd part wrong. Under emergency conditions, the OSM/CRS can grant permission to enter; and the OSM/CRS maintains a set of LHRA Master Keys.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 150

Technical Reference(s): HPP 3104.01, Revision 59 RP-AA-103-1002, Revision 2 Proposed References to be provided to applicants during examination: N Learning Objective: (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments: DAEC is shutdown.

At the request of the Operations Department, HP has constructed a Temporary Locked High Radiation Area (LHRA). It is expected that Operations Department work will be conducted in this area for the next four days.

Subsequently, at the start of the shift, an operator who has been briefed on the job and is on the associated RWP is seeking entry into the Temporary LHRA.

Which ONE of the following identifies...

(1) who can grant the operator permission to enter the Temporary LHRA AND (2) who can issue the key?

A. (1) EITHER HP or the OSM/CRS can grant permission to enter, AND (2) ONLY HP can issue the entry key.

B. (1) ONLY HP can grant permission to enter, AND (2) ONLY HP can issue the entry key.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 151

C. (1) ONLY HP can grant permission to enter, AND (2) EITHER HP or the OSM/CRS can issue the entry key.

D. (1) EITHER HP or the OSM/CRS can grant permission to enter, AND (2) EITHER HP or the OSM/CRS can issue the entry key.

Answer: B Answer Explanation:

ACP-1411.13 (p4, 14; Rev 30)

Steps 3.2 (3) and (5) (a), Attachment 2, Step (2), Bullet 3 Correct - 1st part correct, 2nd part correct. According to ACP-1411.13 (p14; Rev 30) Attachment 2, Step (2), Bullet 3, permission to enter the area may be granted only by a Senior/Journeyman HP technician or HP Supervisor (or a Control Room Supervisor / Operations Shift Manager in an emergency). Since this is NOT an emergency, ONLY the HP representative can grant permission to enter. According to ACP-1411.13 (p4; Rev 30) Step 3.2 (3), the Master keys for LHRA shall be under the administrative control of Health Physics Supervisor and the Operations Shift Manager/Control Room Supervisor. However, according to ACP-1411.13 (p4; Rev 30) Step 3.2 (5) (a) the Operations Shift Manager/Control Room Supervisor on duty shall only be used for urgent or emergency access to these areas as determined by the Operations Shift Manager/Control Room Supervisor. Since this is NOT an emergency, the Key must be obtained from HP.

Answer: B Plausible Distractors:

A. Incorrect - 1st part wrong, 2nd part correct. Under emergency conditions, the OSM/CRS can grant permission to enter.

C. Incorrect - 1st part correct, 2nd part wrong. The OSM/CRS maintains a set of LHRA Master Keys.

D. Incorrect - 1st part wrong, 2nd part wrong. Under emergency conditions, the OSM/CRS can grant permission to enter; and the OSM/CRS maintains a set of LHRA Master Keys.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 152

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 2.4.28 Importance Rating Knowledge of procedures related to a security event (non-safeguards information).

Proposed Question: RO Question # 73 (Question withheld from public disclosure due to containing non-safeguards security-related information.)

A.

B.

C.

D.

Proposed Answer:

Explanation (Optional):

A.

B.

C.

D.

Technical Reference(s):

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 153

Proposed References to be provided to applicants during examination:

Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content:

Comments:

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 154

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 2.4.18 Importance Rating Knowledge of the specific bases for EOPs. CFR: 41.10 Proposed Question: RO Question # 74 The "Torus Level Control Leg" of EOP 2, Primary Containment Control, directs the operators to maintain Torus level above 7.1 feet. In the event Torus water level cannot be maintained, a reactor scram is required.

Which of the following describes the condition this action is intended to prevent?

A. a loss of the pressure suppression function of the Torus by maintaining the Drywell-to-Torus downcomers adequately submerged B. over pressurizing the Torus with HPCI running and exhausting directly to the Torus air space C. over pressurizing the Torus with an SRV open due to uncovering the "T-Quenchers" and bypassing the pressure suppression function D. a loss of Torus level indication by maintaining the lower level instrument tap adequately submerged Proposed Answer: A Explanation (Optional):

A. Correct: EOP Basis states a torus level of 7.1 ft. corresponds to the bottom of the drywell-to-torus downcomers. Torus levels below 7.1 ft. would result in loss of the pressure suppression function of the primary containment (e.g., during a LOCA, steam entering the torus would not be fully condensed).

B. Incorrect: The HPCI Turbine exhaust line discharges at 5.8' torus water level, which must be kept covered, or steam exhaust would not be condensed, threatening containment. 5.8' is close to 7.1', but incorrect.

C. Incorrect: The SRV downcomers discharge at 4.5' Torus water level, which must be kept covered, or steam would not be condensed, threatening containment. 4.5' is close to 7.1 ', but incorrect.

D. Incorrect: One of the Torus level breakpoints is due to the level instrument tap, however the breakpoint is not 7.1'.

Technical Reference(s): Bases-EOP 2, Revision 14 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 155

Proposed References to be provided to applicants during examination: N Learning Objective: 95.60.03.02 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments: The "Torus Level Control Leg" of EOP 2 directs the operators to maintain Torus level above 7.1 feet; and, if it can't be, the reactor shall be scrammed.

The basis of this is to prevent...

A. a loss of the pressure suppression function of the Torus by maintaining the Drywell-to-Torus downcomers adequately submerged.

B. over pressurizing the Torus with HPCI running and exhausting directly to the Torus air space.

C. over pressurizing the Torus with an SRV open due to uncovering the "T-Quenchers" and bypassing the pressure suppression function.

D. a loss of Torus level indication by maintaining the lower level instrument tap adequately submerged.

Answer: A Answer Explanation:

ANSWER:

EOP Basis states a torus level of 7.1 ft. corresponds to the bottom of the drywell-to-torus downcomers. Torus levels below 7.1 ft. would result in loss of the pressure suppression function of the primary containment (e.g., during a LOCA, steam entering the torus would not be fully condensed).

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 156

DISTRACTORS:

The HPCI Turbine exhaust line discharges at 5.8' torus water level, which must be kept covered, or steam exhaust would not be condensed, threatening containment. 5.8' is close to 7.1', but incorrect.

The SRV downcormers discharge at 4.5' Torus water level, which must be kept covered, or steam would not be condensed, threatening containment. 4.5' is close to 7.1 ', but incorrect.

One of the Torus level breakpoints is due to the level instrument tap, however the breakpoint is not 7.1'.

50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 157

Examination Outline Cross-reference: Level RO SRO Tier # 3 Group #

K/A # General 2.4.19 Importance Rating 3.4 Knowledge of EOP layout, symbols, and icons.

Proposed Question: RO Question # 75 While reviewing the EOP flowcharts you come across a symbol that is a DIAMOND SHAPE with an arrow exiting the right side and another arrow out the bottom of the DIAMOND SHAPE.

See below.

What does this symbol indicate?

A. Decision Step B. Hold/Wait Point C. Instructional Step D. Concurrent Execution Proposed Answer: A Explanation (Optional):

A. Correct B. Incorrect This is an octagon C. Incorrect This is a box D. Incorrect this is a downward triangle Bases Flow Chart Use, Rev. 10 pg Technical Reference(s):

12 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 158

Proposed References to be provided to applicants during examination: N Learning Objective: 95.00.00.01 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2003 Fermi 2 Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments: Question #99 on the 2003 Fermi RO Written Exam 50007 Rev. 0 PDA 15-1 RO NRC Written exam 50007_PDA 15-1 RO NRC Written_xm Page 159

WRITTEN / ORAL EXAMINATION KEY Page 1 COVER SHEET Examination Number/Title: PDA 15-1 RO NRC Written Exam / PDA 15-1 SRO NRC Written Exam Training Program: Licensed Operator Initial Training Course/Lesson Plan Number(s): 50007 Reactor Operator / 60006 Senior Reactor Operator PASS CRITERIA:

Total Points Possible: 100 Combined Average 80%. Must Exam Time: 480 minutes Score 70% on SRO Section.

Yes No Yes No This is an alternate examination; verified This is a remediation exam; verified at at least 30% of the questions are least 90% of the questions are different different from other forms/versions of this from the failed exam. For LOIT remedial exam (e.g., Forms A, B, C; continuing exams, verified 95% difference. For training exam versions for consecutive LOCT annual operating and biennial weeks). For LOCT weekly exams during comprehensive remedial exams, verified a segment, verified > 50% difference. no repeat questions.

This is an initial training examination; This is a randomly generated electronic verified at least 30% of the questions are exam printout; verified the exam bank has different from same exam administered 3 questions per objective if one test item to the previous class. on exam for the objective. If 2 or more test items on exam for an objective, then 6 questions are in bank.

Exam development and review guidelines: Key should contain the following:

o TR-AA-230-1003, SAT Development Learning Objective Number Test Item o Question or Statement o All possible answers o Correct Answer Indicated o Point Value o References (if applicable)

NOTE: NRC exams may require additional information. Refer to fleet and site specific procedures.

EXAMINATION REVIEW AND APPROVAL:

Developed by: Date:

Instructional Review of Written Exam (Qualified Instructor): Date:

Technical Review (SME): Date:

Approved by Training Supervisor: Date:

Approved by Training Program Owner (or line designee): Date:

Indicate in the following table if any changes are made to the exam after approval:

AR/TWR# PREPARER DATE

DESCRIPTION OF CHANGE REASON FOR CHANGE (if applicable) SUPERVISOR DATE Filename: 60006_PDA 15-1 RO-SRO NRC Written_xm TR-AA-230-1003-F13 Revision 1

WRITTEN / ORAL EXAMINATION COVER SHEET Page 1 Trainee Name:

Employee Number: Site: DAEC Examination Number/Title: PDA 15-1 RO NRC Written Exam / PDA 15-1 SRO NRC Written Exam Training Program: Licensed Operator Initial Training Course/Lesson Plan Number(s): 50007 Reactor Operator / 60006 Senior Reactor Operator RO Section Grade:

RO Total Points: 75 PASS CRITERIA: /75= %

SRO Total Points: 25 Combined Average 80%. Must Score 70% on SRO Section. SRO Section Grade:

Total Points Possible: 100

/25= %

Graded by: Date:

Co-graded by (if necessary): Date:

EXAMINATION RULES

1. References may not be used during this examination, unless otherwise stated.

2. Read each question carefully before answering. If you have any questions or need clarification during the examination, contact the examination proctor.

3. Conversation with other trainees during the examination is prohibited.

4. Partial credit will not be considered, unless otherwise stated. Show all work and state all assumptions when partial credit may be given.

5. Rest room trips are limited and only one examinee at a time may leave.

6. For exams with time limits, you have 480 minutes to complete the examination.

7. Feedback on this exam may be documented on TR-AA-230-1004-F03, Examination Feedback Form. Contact Instructor to obtain a copy of the form.

EXAMINATION INTEGRITY STATEMENT Cheating or compromising the exam will result in disciplinary actions up to and including termination.

I acknowledge that I am aware of the Examination Rules stated above. Further, I have not given, received, or observed any aid or information regarding this examination prior to or during its administration that could compromise this examination.

Examinees Signature: Date:

REVIEW ACKNOWLEDGEMENT I acknowledge that the correct answers to the exam questions were indicated to me following the completion of the exam. I have had the opportunity to review the examination questions with the instructor to ensure my understanding.

Examinees Signature: Date:

Filename: 60006_PDA 15-1 RO-SRO NRC Written_xm TR-AA-230-1003-F12 Revision 1

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A 295001 AA 2.06 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION : Nuclear boiler instrumentation. (CFR: 41.10 / 43.5)

Proposed Question: SRO Question # 76 The plant was operating at 100% reactor power when reactor recirculation pump A tripped, resulting in the following indications:

Reactor Recirculation Loop A Jet Pump Flow is 2 Mlbm/hr

Reactor Recirculation Loop B Jet Pump Flow is 26 Mlbm/hr

Core Plate P as indicated on PDR-4528 is 4.4 psid

Reactor Power is 65% Rated Thermal Power (RTP)

Determine which of the following actions should be directed.

A. Insert control rods until Reactor Power is less than < 60% RTP - ONLY B. Reduce Reactor Recirculation Pump B speed until Reactor Power is less than 60%

RTP - ONLY C. Reduce Reactor Recirculation Pump B speed until Reactor Recirculation Loop B Jet Pump Flow is less than 25 Mlbm/hr - ONLY D. Reduce Reactor Recirculation Pump B speed until both Reactor Power is less than 60% RTP AND Reactor Recirculation Loop B Jet Pump Flow is less than 25 Mlbm/hr Proposed Answer: A Explanation (Optional):

A. Correct - Per the administrative limits given AOP 264, Reactor Power shall be less than or equal to 60% RTP and the procedure directs that the power reduction be accomplished using control rods.

B. Incorrect - AOP 264 directs that the power reduction be accomplished using control rods. Plausible since reducing flow will also reduce power.

C. Incorrect - Per the administrative limits given AOP 264, Total Core Flow shall be maintained less than or equal to 25.95 Mlbm/hr. Total Core flow is 25 Mlbm/hr (Loop B flow minus Loop A flow and can be verified using the Core Flow vs Core Plate d/p).

Plausible if applicant believes that loop flow must be maintained less that the limit.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 2

D. Incorrect - See above.

AOP 255.2, Power/Reactivity Technical Reference(s): AOP 264, Loss of Recirc Pump(s)

Abnormal Change IPOI 3, Power Operations (33% -

100% Rated Power)

Core Flow vs Core Plate Differential Pressure Proposed References to be provided to applicants during examination:

DAEC Power/Flow Map Learning Objective: 5.58.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

SRO Only Guidance E. Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations. [10 CFR 55.43(b)(5)]

This 10 CFR 55.43 topic involves both 1) assessing plant conditions (normal, abnormal, or emergency) and then 2) selecting a procedure or section of a procedure to mitigate, recover, or with which to proceed.

One area of SRO level knowledge (with respect to selecting a procedure) is knowledge of the content of the procedure versus knowledge of the procedures overall mitigative strategy or purpose.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 3

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A # 295004.A2.03: Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF D.C. POWER: Battery voltage.

Proposed Question: SRO Question # 77 Given the following:

The plant is operating at 75% power

During routine checks on 125VDC Battery 1D1 it is observed that pilot cell electrolyte level is below the minimum level indication mark, but remains above the top of the plates

Electrical Maintenance subsequently reports that battery pilot cell float voltage for 125VDC Battery 1D1 is 2.08.V Based upon the conditions above, TS 3.8.6 Condition(s) __(1)__ is (are) applicable, and the Division I 125 VDC Battery can be considered ______(2)_______.

A. (1) A ONLY (2) OPERABLE B. (1) A ONLY (2) INOPERABLE C. (1) A AND B (2) OPERABLE D. (1) A AND B (2) INOPERABLE Proposed Answer: A Explanation (Optional):

A. Correct.

B. Incorrect: Low electrolyte level (Categories A and B) will result in entry in ITS 3.8.6 Condition A. Condition A requires checking battery float voltage; the values provided in the stem for electrolyte level and battery voltage meet the minimum requirement of Category C and will not result in additional entry into ITS 3.8.6 Condition B (Condition B directs that the affected battery must be declared inoperable). The second part of the question does not refer to the battery operability directly, but rather the Division I 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 4

125 VDC distribution system as a whole. To address operability at that level, knowledge of the basis of ITS 3.8.6 is required; specifically the basis for Condition B states that the corresponding DC electrical power subsystem must be declared inoperable. This is also part of the ITS 3.8.4 basis which requires that the associated battery be operable for the DC subsystem to be operable. In this instance, operability is maintained.

C. Incorrect. Low electrolyte level (Categories A and B) will result in entry in ITS 3.8.6 Condition A. Condition A requires checking battery float voltage; the values provided in the stem for electrolyte level and battery voltage meet the minimum requirement of Category C and will not result in additional entry into ITS 3.8.6 Condition B (Condition B directs that the affected battery must be declared inoperable). The second part of the question does not refer to the battery operability directly, but rather the Division I 125 VDC distribution system as a whole. To address operability at that level, knowledge of the basis of ITS 3.8.6 is required; specifically the basis for Condition B states that the corresponding DC electrical power subsystem must be declared inoperable. This is also part of the ITS 3.8.4 basis which requires that the associated battery be operable for the DC subsystem to be operable. In this instance, operability is maintained.

D. Incorrect: Low electrolyte level (Categories A and B) will result in entry in ITS 3.8.6 Condition A. Condition A requires checking battery float voltage; the values provided in the stem for electrolyte level and battery voltage meet the minimum requirement of Category C and will not result in additional entry into ITS 3.8.6 Condition B (Condition B directs that the affected battery must be declared inoperable). The second part of the question does not refer to the battery operability directly, but rather the Division I 125 VDC distribution system as a whole. To address operability at that level, knowledge of the basis of ITS 3.8.6 is required; specifically the basis for Condition B states that the corresponding DC electrical power subsystem must be declared inoperable. This is also part of the ITS 3.8.4 basis which requires that the associated battery be operable for the DC subsystem to be operable. In this instance, operability is maintained.

DAEC Technical Specifications Technical Reference(s):

3.8.4 and 3.8.6 DAEC Technical Specifications Bases 3.8.4 and 3.8.6 DAEC ITS 3.8.6 to include Table 3.8.6-Proposed References to be provided to applicants during examination: 1 (no basis information to be provided)

Learning Objective: 1.02.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 5

Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 6

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295018 2.1.20 Importance Rating Partial or Total Loss of CCW: Ability to interpret and execute procedure steps. CFR: 41.10 /

43.5 Proposed Question: SRO Question # 78 The plant was operating at 100% reactor power when the operating GSW pumps tripped. The following energized:

1C06A (C-3), "A" GSW Pump 1P 89A Trip OR Motor Overload

1C06A (C-4), "B" GSW Pump 1P 89B Trip OR Motor Overload

1C06A (B-3), GSW Pumps 1P 89A/B/C DISCH Header LO Pressure The operators entered and began executing the steps of AOP 411, GSW Abnormal Operation.

The operators were unsuccessful in starting ANY GSW pumps and the following energized:

1C05A (A-8), PCIS Channel "A" Steam Tunnel HI TEMP

1C05B (A-7), PCIS Channel "B" Steam Tunnel HI TEMP Shortly thereafter, the MSIVs shut. The CRS should direct the operators to:

A. Perform a Fast Power Reduction then insert a manual scram ONLY B. Enter EOP-3, Secondary Containment Control, and perform IPOI 5, Reactor Scram, actions C. Trip both Recirc MG Sets then insert a manual scram D. Send an operator to verify both Steam Tunnel Cooling fans 1V AC 17A and B are running Proposed Answer: B Explanation (Optional):

A. Incorrect - AOP 411 states that if no GSW pumps are running and flow cannot be restored immediately then manually scram the reactor. A fast power reduction is not performed with the conditions provided in the STEM.

B. Correct - AOP 411, Step 7 (p. 4), states if the MSIVs close on steam tunnel high temperature then enter EOP-3 due to the area Max Normal Temperature being exceeded and verify reactor scrammed.

C. Incorrect - required if high temperatures occur on the recirc MG sets.

D. Incorrect - would have been completed prior to the MSIV closure.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 7

Technical Reference(s): AOP 411, Revision 27 Proposed References to be provided to applicants during examination: N Learning Objective: 6.67.01.06 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 8

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A #: Generic K/A 2.2.38: Knowledge of conditions and limitations in the facility license.

Associated topic: 295028 (High Drywell Temperature)

Proposed Question: SRO Question # 79 Given the following:

Analysis indicates that maximum allowed drywell temperature limit of Technical Specification 3.6.1.4, Drywell Air Temperature, can be raised to 136°F Which of the following describes the correct process for making the change described above?

A. The plant may update the technical specification, but must subsequently inform the NRC in a biannual report B. The plant must transmit the revised technical specification to the NRC in accordance with TS 5.5.10 (Technical Specification Bases Control Program)

C. The plant must file an application for a technical specification amendment with the NRC prior to making the change D. The plant may make technical specification changes that are fully bounded by the Safety Evaluation without informing the NRC Proposed Answer: C Explanation (Optional):

A. Incorrect: The plant makes a biennial update to the NRC for UFSAR changes, not for TS (ACP 102.24 page 7 and 10 CFR 50.59).

B. Incorrect: After implementation of TS Bases changes the revised Bases pages are formally transmitted to the NRC in accordance with the Technical Specification Bases Control Program (TS 5.5.10). This program only concerns the Bases however, and not TS (ACP 102.24 page 7 and DAEC TS 5.5.10).

C. Correct: when a licensee desires to amend their license (of which TS are a part),

application for an amendment must be filed with the NRC (10 CFR 50.90).

D. Incorrect: when processing TS Bases changes associated with implementation of a licensing amendment, plant staff ensures that those changes are fully bounded by the 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 9

NRC safety evaluation for the amendment. This scenario involves a change to the TS themselves however, and not just the TS Bases (ACP 102.24 page 5).

10 CFR 50.59 Changes, Tests and DAEC TS 3.6.1.4, Drywell Air Technical Reference(s):

Experiments. Temperature 10 CFR 50.90 Application for DAEC TS 5.5.10, Technical Amendment of License, Specifications (TS) Bases Control Construction Permit, or Early Site Program Permit.

ACP 102.24, Preparation, Review, and Processing of Bases Changes, Revision 9 Proposed References to be provided to applicants during examination: None Learning Objective: 1.02.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 X Comments:

Comments: SRO-only question justification is the link to 10CFR55.43(b)(1) Conditions and limitations in the facility license.

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question covers Processes for TS and FSAR changes.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 10

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295038 2.2.44 Importance Rating Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions. (CFR: 41.5

/ 43.5 / 45.12)

Proposed Question: SRO Question # 80 A radiological release occurred while operating at power.

Annunciator 1C35A, C-3 Reactor BLDG KAMAN 3,4,5,6,7 & 8 HI RAD OR Monitor Trouble, was activated

Annunciator 1C05B, C-8 PCIS Group 3 Isolation Initiated, was activated

Reactor Building Ventilation has isolated

Both Standby Gas Treatment (SBGT) trains are operating

1C23, Reactor Building to atmosphere indicates -1.1 inches water

Turbine Building Ventilation has shutdown

Offsite release is above the ALERT Level In accordance with EOP-4, Radioactivity Release Control, which ventilation system would the CRS direct re-started and why?

A. Turbine Building Ventilation to filter ventilation exhaust from the Turbine Building.

B. Main Plant Exhaust Fans to prevent unmonitored ground release of radioactivity.

C. Turbine Building Ventilation to prevent unmonitored ground release of radioactivity.

D. Reactor Building Ventilation to reduce the Reactor Building area and equipment temperatures Proposed Answer: C Explanation (Optional):

A. Incorrect: There is no filtration on TB exhaust (EOP-4 Bases).

B. Incorrect: Common misconception, EF 1,2, & 3 do not trip on Group 3. Their exhaust from the plant is the sample point for Kaman 3-8. Main Plant exhaust fans are tripped on a RB Kaman Hi Hi alarm concurrent with Group III isolation to prevent bypass of the SGTS filter units by air from the RB via main plant ventilation stack.

C. Correct.

D. Incorrect: RB vents are isolated and exhaust is routed to the main plant exhaust plenum. On a Group III isolation the Main Plant Exhaust fans are secured to prevent bypassing the SGTS filter units to preclude or limit untreated release to the environs 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 11

(EOP-4 Bases).

Technical Reference(s): EOP Bases, Rev. 9, pages 4 - 5 ARP 1C05B, C-8, Rev. 98 ARP 1C35A, C-3, Rev. 43 Proposed References to be provided to applicants during examination: N Learning Objective: 6.72.02.03 (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam: 2005 Fitzpatrick Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 12

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 600000 2.4.35 Importance Rating (Plant Fire On Site) Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. (CFR: 41.10 / 43.5)

Proposed Question: SRO Question # 81 Given the following:

The Main Control Room has been evacuated due to a fire

Torus Water Temperature is 106°F Where can Torus Water Temperature be obtained, and what action will the CRS direct to lower Torus Water Temperature?

A. Remote Shutdown Panel, 1C-392; it is required to maximize Torus Cooling with B RHR Loop IAW OI-149, Residual Heat Removal System.

B. Remote Shutdown Panel, 1C-392; it is required to maximize Torus Cooling with B RHR Loop IAW AOP-915, Shutdown Outside Control Room.

C. Remote Shutdown Panel, 1C-388; it is required to maximize Torus Cooling with B RHR Loop IAW OI-149, Residual Heat Removal System.

D. Remote Shutdown Panel, 1C-388; it is required to maximize Torus Cooling with B RHR Loop IAW AOP-915, Shutdown Outside Control Room.

Proposed Answer: D Explanation (Optional):

A. Incorrect: is plausible-location of TRANSFER Switch for TI-4325A (not indication), and wrong procedure for Torus Cooling operation.

B. Incorrect: is plausible-location TRANSFER Switch for TI-4325A (not indication).

C. Incorrect: is plausible-location of indication is correct, except wrong procedure for Torus Cooling operation.

D. Correct: location of indication is correct, and Torus Cooling operation is directed by AOP-915 per Section 4.

Technical Reference(s): EOP 2, Rev. 16 AOP 915, Rev. 53 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 13

Proposed References to be provided to applicants during examination: N Learning Objective: 6.62.01.01 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 14

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating Proposed Question: SRO Question # 82 Given the following:

The plant is operating at 100% power

Lightning strikes are occurring in the vicinity of the plant site

ITC MIDWEST has reported minor grid fluctuations as a storm passes through the region

Main Generator MVAR output spikes to 270 MVARS (out) and remains there

Main Generator hydrogen gas pressure is currently 45 psig, and megawatt output is 640 MWe Which of the following procedure sections should be implemented immediately?

A. AOP-304, Grid Instability, section titled Preparation for High Grid Loading and Potential Instability B. AOP-304, Grid Instability, section titled Grid Instability C. AOP-903, Severe Weather, section titled High Wind / Severe Weather / Tornado Watch

[Advisory]

D. AOP-903, Severe Weather, section titled High Wind / Severe Thunderstorm Warning Proposed Answer: B Explanation (Optional):

A. Incorrect: AOP-304s PREPARATION FOR HIGH GRID LOADING AND POTENTIAL INSTABILITY section could possibly be implemented based upon the stem conditions; however it only addresses ensuring that the generator voltage regulator is in automatic.

It does not contain actions to control generator output if the generator capability curve is challenged (AOP-304 page 4).

B. Correct: MVAR loading is exceeding the generator capability curve based upon the conditions in the stem (OI-698 page 45). AOP-304s GRID INSTABILITY section contains steps to: Establish critical parameter monitoring of Main Generator MVARs.

Take actions as directed by ARPs and OIs and reduce reactor power and generator output as necessary to comply with DAEC procedures and protect DAEC equipment even if this will result in further degradation of the grid as necessary to maintain 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 15

equipment within operating specifications. Monitor generator parameters on the Generator Estimated Capability Curve. Return Generator Voltage and MVARs to the desired level, once the grid has stabilized and as allowed by ITC MIDWEST (AOP-304 pages 8-12)

C. Incorrect: AOP-903s HIGH WIND / SEVERE WEATHER / TORNADO WATCH

[ADVISORY] section could possibly be implemented based upon the conditions in the stem, but does not provide direction regarding response to storm related electrical effects to the grid/plant (AOP-903 pages 4-6).

D. Incorrect: AOP-903s HIGH WIND / SEVERE THUNDERSTORM WARNING section could possibly be implemented given that the storm has arrived at the site in the stem conditions, however it only contains steps to: Contact ITC to check status of grid stability. Relay information to the Real Time Desk and ITC regarding any known disturbances within the plant distribution network, and any plant malfunctions that may increase the potential for disturbing plant electrical output during a Severe Thunderstorm, or High Wind condition (AOP-903 pages 7-9)

Technical Reference(s): AOP 304, Revision 40 AOP 903, Revision 49 OI-698, Revision 87 OI-698, Appendix 1, Proposed References to be provided to applicants during examination: Estimated Capability Curves Learning Objective: 5.48.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments: Comments: SRO-only question justification is the link to 10CFR55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question covers Assessing plant conditions (normal, abnormal, or emergency) and then selecting a 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 16

procedure or section of a procedure to mitigate, recover, or with which to proceed.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 17

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295014 G 2.1.23 Importance Rating (Inadvertent Reactivity Addition) Ability to perform specific system and integrated plant procedures during all modes of plant operation. CFR: 41.10 / 43.5 Proposed Question: SRO Question # 83 DAEC is performing a startup from a planned outage. The following plant conditions are given:

Reactor Power is 6%

Reactor Pressure is 770 psig.

Control rod withdrawal is in progress

The last reactor startup occurred 400 days ago A Rod Worth Minimizer self-test failure occurs under these conditions.

Can the control rod withdrawal continue and what is the basis for this decision?

A. No, the only control rod movement allowed is by reactor scram B. Yes, the RWM may be bypassed as long as a qualified member of plant staff verifies control rod movement C. No, the only control rod movement allowed is rod insertion using Emergency In D. Yes, the RWM is Auto Bypassed when APRM power is >5% and RWM restrictions are not enforced at this power level Proposed Answer: B Explanation (Optional):

A. Incorrect: Although the RWM does prevent rod movement in its current configuration, the RWM may be bypassed and control rod movement may continue.

B. Correct: It is permissible to bypass the RWM and utilize a member of the plant staff to verify control rods are within the BPWS provided no other startup has been conducted within the last calendar year with the RWM inoperable. 400 days was provided in the stem.

C. Incorrect: Although the RWM does prevent rod movement in its current configuration, the RWM may be bypassed and control rod movement may continue.

D. Incorrect: Although the RWM does prevent rod movement in its current configuration, the RWM may be bypassed and control rod movement may continue.

Technical Reference(s): TS 3.3.2.1 Condition C 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 18

TS 2.0, Amend. 243 10CFR50.36 Proposed References to be provided to applicants during examination: TS 3.3.2.1 Learning Objective: 1.02.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments: SRO question per 10 CFR 55.43.1 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 19

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A# 295015.A2.01: Ability to determine and/or interpret the following as they apply to INCOMPLETE SCRAM: Reactor power.

Proposed Question: SRO Question # 84 Given that the plant was operating at approximately full power with the following conditions:

A failed FULL-IN reed switch on ONE control rod

A complete loss of 1Y23, 120V Uninterruptable AC Distribution, then occurred

The reactor scrammed EOP 1, RPV Control, was entered due to RPV Low Level during the initial transient:

All 8 RPS Scram Group A and B white lights were OFF

The Operator at the controls could not confirm that all rods were fully inserted

On the 1C05 Full Core Display, all LPRM downscale lights were ON

All IRMs were fully inserted, on range 3 or 4, reading mid-scale, and lowering on all available indications

RPV pressure was 900 psig and lowering slowly with all available Main Steam Lines Drains open

Standby Liquid Control (SBLC) had NOT injected

There were no challenges to containment Which of the following correctly describes the correct procedure usage when directing further operator actions in this situation?

A. ALL operator actions must be directed from EOP 1 and IPOI 5. NO operator actions should be directed from the ATWS EOP B. Operator actions for reactivity control must be directed from the ATWS EOP. Operator actions for RPV level and pressure must be directed from EOP 1 C. Operator actions for reactivity control will be directed from IPOI 5. Operator actions for RPV Pressure and Level must be directed from the ATWS EOP D. NO operator actions should be directed from either EOP 1 or IPOI 5. ALL operator actions must be directed from the ATWS EOP Proposed Answer: C Explanation (Optional):

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 20

A. Incorrect: Selected if Reactor is believed to be SD under all conditions per EOP 1 1st Recheck Statement. IPOI-5 is performed concurrently per EOP 1 RC-2.

B. Incorrect: Selected if ATWS is entered but the /Q 1st Recheck Statement is misapplied.

C. Correct: Rod position indication is lost with a UPS failure, which would also result in a Reactor Scram. The 1C05 APRM/IRM recorders are powered from Inst AC on the "A" channel and UPS on the "B" channels. Embedded in this question is the definition of "SHUTDOWN" from EOP Bases Flowchart Use and Logic: Reactor subcritical (power decreasing), and below point of adding heat (POAH) which is 20 on IRM Range 8. IRMs are on Range 3-4 and lowering. Reactor is not SD under all conditions so ATWS must be entered and /1 & /2 performed. At the top of the /Q leg is a Recheck Statement that says exit this flowpath (exit /Q only) and reenter IPOI-5 if reactor is shutdown. CRS must make this operational judgment or the next steps in /Q will trip the Recirc Pumps.

CRS should remain in ATWS /L and /P legs. This question is based on a plant event.

D. Incorrect: Selected if the /Q 1st Recheck Statement is not observed or the definition of shutdown is not understood.

Technical Reference(s): EOP 1, Revision 18 AOP 357, Revision 44 ATWS, Revision 21 BASES ATWS, Revision 17 Proposed References to be provided to applicants during examination: None Learning Objective: 6.55.01 (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New MODIFIED FROM Question History: Last NRC Exam:

BANK (DAEC)

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

Comments: SRO-only question justification is the link to 10CFR55.43(b)(5) Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 21

covers Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 22

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 295036 G 2.4.49 Importance Rating (Secondary Containment High Sump/Area Water Level/5) Ability to perform without reference to procedures those actions that require immediate operation of system components and controls. CFR: 41.10 / 43.2 Proposed Question: SRO Question # 85 With the plant operating at 100% power, a fire was discovered in the Reactor Building.

The fire was extinguished with the deluge system, resulting in these conditions:

HPCI Room Temperature is 160°F

HPCI Room Water Level is 3 inches, (Max Normal Level = 2 inches)

Which of the following actions is required?

A. Enter IPOI-5, Reactor Scram and perform a subsequent plant shutdown.

B. Enter Emergency Depressurization and open Safety Relief Valves.

C. Enter EOP-3, Secondary Containment Control, and isolate the deluge system.

D. Enter EOP-1, RPV Control, and anticipate Emergency Depressurization.

Proposed Answer: C Explanation (Optional):

A. Incorrect: a reactor shutdown per the IPOIs would be performed if the same parameter was above Max Safe in two areas and a reduction in reactor pressure would not affect the leak rate.

B. Incorrect: an ED would be performed it two areas were above Max Safe and if reduction in reactor pressure would affect the leak rate.

C. Correct: per EOP 3, when any parameter is above Max Normal (HPCI room water level of 2 inches is at Max Normal) and the system is not required per the EOPs or to suppress a fire.

D. Incorrect: if any parameter is above Max Safe and a reduction in reactor pressure would affect the leak rate.

Technical Reference(s): Bases-EOP 3, Rev 20 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 23

Proposed References to be provided to applicants during examination: EOP 3, Table 6 Learning Objective: 6.67.10.01 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 24

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 205000 A2.02 Importance Rating Ability to (a) predict the impacts of the following on the SHUTDOWN COOLING SYSTEM (RHR SHUTDOWN COOLING MODE); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low shutdown cooling suction pressure.

Proposed Question: SRO Question # 86 Given the following:

The plant is in Mode 4

The Residual Heat Removal (RHR) system is aligned for shutdown cooling

RHR Pump "A" is running

An equipment failure results in a spurious Primary Containment Isolation System (PCIS) Group 4 full isolation Based upon these conditions, which of the following describes the correct strategy to restore shutdown cooling?

A. Use AOP-149, Loss of Decay Heat Removal, to TRIP the "A" RHR pump and subsequently restore shutdown cooling B. Use ARP 1C05B(D-8), PCIS Group "4" Isolation Initiated, to restore Group 4 Isolation valves to the desired lineup, and restore shutdown cooling C. Use AOP-149, Loss of Decay Heat Removal, to manually OPEN shutdown cooling suction valves MO-1908 and MO-1909 and restore shutdown cooling D. Use OI-264, Reactor Recirculation System, to START a Recirculation Pump to provide forced circulation Proposed Answer: C Explanation (Optional):

A. Incorrect: the spurious Group 4 isolation causes shutdown cooling suction valves MO-1908 and MO-1909 to both go shut. This removes the only suction pathway for RHR Pump A. To prevent damage to RHR Pump A due to low suction pressure, a protective feature will cause an automatic trip. The logic of this protective feature is such that as soon either MO-1908 or MO-1909 begin to close, a trip of RHR Pump A would immediately result (SD-149 page 12). While it would be reasonable to take action stop a pump which has lost suction in order to prevent damage, this action would not be necessary since it would occur automatically. Additionally, this action is 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 25

contained in AOP-149.

B. Incorrect: ARP 1C05B (D-8) contains steps to correct the cause of the Group 4 isolation, reset the Group 4 logic, and to return Group 4 logic to the desired lineup.

Prior to that point in the procedure however, ARP 1C05B (D-8) provides direction to perform AOP-149 if RHR was operating in the shutdown cooling mode. AOP-149 will subsequently provide direction that if an equipment failure results in an invalid Group 4 isolation, cooling may be restored by manually opening the applicable isolation valve (ARP 1C05B, AOP-149 pages 4 - 5).

C. Correct: the spurious Group 4 isolation causes shutdown cooling suction valves MO-1908 and MO-1909 to both go shut. This removes the only suction pathway for RHR Pump A. To prevent damage to RHR Pump A due to low suction pressure, a protective feature will cause an automatic trip. The logic of this protective feature is such that as soon either MO-1908 or MO-1909 begin to close, a trip of RHR Pump A would immediately result (SD-149 page 12). AOP-149 contains direction that if an equipment failure results in an invalid Group 4 isolation, cooling may be restored by manually opening the applicable isolation valve (AOP-149 pages 4 - 5).

D. Incorrect: AOP-149 directs the starting of a Recirc Pump using OI-264 if shutdown cooling cannot be restored. Prior to that point in the procedure however, AOP-149 provides direction that if an equipment failure results in an invalid Group 4 isolation, cooling may be restored by manually opening the applicable isolation valve (AOP-149 pages 4 - 6).

Technical Reference(s): AOP 149, Revision 15 ARP 1C05B, Revision 98 SD-149, Revision 13 Proposed References to be provided to applicants during examination: None Learning Objective: (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 26

Comments: Comments: SRO-only question justification is the link to 10CFR55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question covers 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.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 27

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A#: Generic K/A 2.4.18: Knowledge of the specific bases for EOPs. Associated topic:

211000 (SLC).

Proposed Question: SRO Question # 87 Given the following:

The plant was operating at 100% reactor power

A transient occurred and the reactor was manually scrammed The following are current plant conditions:

69 control rods stuck full out

Reactor power is 25%

Defeat 15 is installed

Actions of the ATWS QRC have been completed

MSIVs are open

Torus temperature 72°F and stable

Reactor pressure 880 psig and controlled by EHC

Reactor level is +10 inches and stable

SBLC has NOT been injected The Operator at the Controls reports that reactor power has started to oscillate from 15% to 42% power on all APRMs.

Given these conditions, what actions must be directed with respect to SBLC?

SBLC injection .

A. is not required at this time B. must be injected immediately due to the containment being challenged C. must be injected immediately due to the potential of a PCIS Group 1 isolation on RPV low pressure D. must be injected immediately because power oscillations of this magnitude may damage the fuel clad barrier Proposed Answer: D 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 28

Explanation (Optional):

A. Incorrect - SBLC is required to be injected, as directed by ATWS.

B. Incorrect - SBLC is required to be injected, however there are no challenges to containment from the conditions provided in the STEM.

C. Incorrect - With conditions provided in the STEM, the ATWS QRC will have the operator place the MODE switch in Shutdown. MSIVs will not close on RPV low pressure.

D. Correct - The threshold is used to establish a requirement for boron injection following a failure to scram, thereby terminating flux oscillations and minimizing the potnential for fuel damage. The threshold of 25% is readily observable on neutron monitors and is not expected to threaten cladding integrity.

Technical Reference(s): ATWS, Revision 21 Bases-ATWS, Revision 17 Bases-Curves, Rev. 13 Proposed References to be provided to applicants during examination: None Learning Objective: 6.56.05.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments: Comments: SRO-only question justification is the link to 10CFR55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question covers 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.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 29

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 30

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 212000 A2.15 Importance Rating Ability to (a) predict the impacts of the following on the REACTOR PROTECTION SYSTEM ;

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Load rejection.

Proposed Question: SRO Question # 88 Given the following plant conditions:

Reactor Power was 27% as sensed by turbine 1st stage pressure

While performing Main Turbine Stop valve testing, an operator inadvertently closed Turbine Stop valve "1" while Turbine Stop valve "2" was full closed (1) Which one of the following correctly describes the response of the Reactor Protection System?

AND (2) What procedure should be directed as a result of these conditions?

A. (1) A half scram signal on RPS A is generated (2) Enter and direct the actions of ARP 1C05(A-2), "A" RPS AUTO SCRAM B. (1) Neither a full nor a half scram is generated because the scram signal is bypassed (2) No procedure entry is required C. (1) A full reactor scram is generated because two of four turbine stop valves are fully closed (2) Enter and execute the actions of EOP 1, RPV Control and IPOI 5, Reactor Scram D. (1) A half scram signal on RPS A is generated (2) Enter and execute the steps of IPOI 5, Reactor Scram Proposed Answer: Original answer was 'A'. Answer changed by NRC to 'D' based on post-exam comment. Based on the plant conditions in the stem, an EOC-RPT trip of both Recirculation Pumps would occur. This would in turn require a manual scram, and IPOI-5 would performed.

Explanation (Optional):

A. Correct: Power is high enough to cause RPS channels to trip, but not RPS B. Only the ARP is required to be entered.

B. Incorrect. RPS A will trip.

C. Incorrect. Only RPS A will trip.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 31

D. Incorrect. Wrong procedure Technical Reference(s): ARP 1C05 A-2 IPOI-5 EOP-1 Proposed References to be provided to applicants during examination: N Learning Objective: 5.11.01.02 (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam: 1999 Clinton Power Station Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 32

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A#: Generic K/A 2.2.40: Ability to apply Technical Specifications for a system. Associated system: 215005 (APRM / LPRM).

Proposed Question: SRO Question # 89 Given the following:

The Reactor Mode Switch is in the "Startup/Hot Standby Position"

A and D APRMs are in BYPASS NOTE: The table below contains a summary of Local Power Range Monitor (LPRM) instrumentation. Instruments that are shown bolded with a strikethrough are currently BYPASSED.

APRM-A APRM-B APRM-C APRM-D APRM-E LPRM-B APRM-F LPRM-A Level A 3A-32-33 2A-08-33 3A-24-25 1A-16-41 2A-16-33 3A-16-25 3A-24-33 1A-24-41 4A-16-17 4A-24-17 4A-08-09 4A-32-25 5A-32-17 6A-32-09 4A-08-17 2A-08-25 5A-16-09 5A-40-17 3A-40-25 4A-24-09 Level B 1B-24-41 3B-16-25 3B-32-33 2B-08-33 3B-24-25 3B-24-33 1B-16-41 2B-16-33 2B-08-25 6B-32-09 4B-16-17 4B-24-17 4B-08-09 4B-08-17 4B-32-25 5B-32-17 3B-40-25 5B-40-17 5B-16-09 4B-24-09 Level C 2C-16-33 3C-24-33 1C-24-41 3C-16-25 3C-32-33 1C-16-41 2C-08-33 3C-24-25 5C-32-17 4C-08-17 2C-08-25 6C-32-09 4C-16-17 4C-32-25 4C-24-17 4C-08-09 5C-40-17 3C-40-25 5C-16-09 4C-24-09 Level D 3D-24-25 1D-16-41 2D-16-33 3D-24-33 1D-24-41 2D-08-33 3D-16-25 3D-32-33 4D-08-09 4D-32-25 5D-32-17 4D-08-17 2D-08-25 4D-24-17 6D-32-09 4D-16-17 5D-16-09 5D-40-17 3D-40-25 4D-24-09 "A" RPS "B" RPS "A" RPS "B" RPS "A" RPS "A" RPS "B" RPS "B" RPS Which ONE of the following correctly describes the status of APRM "E", AND whether the channel is currently required per Technical Specification (TS) 3.3.1.1, Reactor Protection System Instrumentation?

A. OPERABLE; NOT currently required by TS 3.3.1.1 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 33

B. OPERABLE; currently required by TS 3.3.1.1 C. INOPERABLE; NOT currently required by TS 3.3.1.1 D. INOPERABLE; currently required by TS 3.3.1.1 Proposed Answer: D Explanation (Optional):

A. Incorrect: the APRM is inoperable because it has fewer than 13 LPRM inputs (APRM 'E' shares LPRM inputs with LPRM 'B') (DAEC TS Bases 3.3.1.1 Item 2.d and OI-878.4 pages 3 - 4). ITS 3.3.1.1 Item 2.d applies in Modes 1 and 2; the stem conditions indicate that the plant is in Mode 2 (DAEC TS 3.3.1.1).

B. Incorrect: the APRM is inoperable because it has fewer than 13 LPRM inputs (APRM 'E' shares LPRM inputs with LPRM 'B') (DAEC TS Bases 3.3.1.1 Item 2.d and OI-878.4 pages 3 - 4).

C. Incorrect: ITS 3.3.1.1 Item 2.d applies in Modes 1 and 2; the stem conditions indicate that the plant is in Mode 2 (DAEC TS 3.3.1.1).

D. Correct.

Technical Reference(s): TS 3.3.1.1, Amend. 223 TS Bases 3.3.1.1, Amend. 223 OI 878.4, Revision 40 Proposed References to be provided to applicants during examination: None Learning Objective: 1.02.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments: Comments: SRO-only question justification is the link to 10CFR55.43(b)(2) 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 34

Facility operating limitations in the Technical Specifications and their bases.

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question covers Knowledge of TS bases that is required to analyze TS required actions and terminology."

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 35

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 217000 A2.01 Importance Rating Ability to (a) predict the impacts of the following on the REACTOR CORE ISOLATION COOLING SYSTEM (RCIC); and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: System initiation signal.

Proposed Question: SRO Question # 90 Given the following:

All control rods are fully inserted

High Pressure Coolant Injection (HPCI) suction swapped to the Torus

HPCI and Reactor Isolation Core Cooling (RCIC) tripped on high Reactor Pressure Vessel (RPV) level

RPV level is 190 inches and stable being controlled with condensate and feed

RPV pressure is 1060 psig and slowly rising

Drywell pressure is 1.5 psig

RCIC Equipment Room temperature is currently 150°F Based upon the conditions above, which Emergency Operating Procedure (EOP) Defeat will be directed by the CRS to use RCIC for RPV pressure control in CST-CST mode?

A. DEFEAT 1, RCIC Low RPV Pressure Isolation and 211 Inches Defeat B. DEFEAT 2, HPCI High Torus Water Level Transfer Defeat C. DEFEAT 8, RCIC Steam Line Isolation Defeat D. DEFEAT 18, HPCI/RCIC Area High TEMP Isolation Defeat Proposed Answer: B Explanation (Optional):

A. Incorrect: The purpose of Defeat 1 is to permit RCIC Steam Isolation valves to be opened, or to remain open, when RPV Pressure is 50 psig or less and to remove the 211" RCIC shutdown signal from the RCIC Turbine Steam Supply Valve. Defeat 1 is authorized for an alternate depressurization system utilized in Emergency Depressurization.

B. Correct. For RPV pressure control purposes, bypassing the HPCI high torus water suction swap with Defeat 2 allows RCIC to be used in the CST- to-CST lineup, since redundant shutoff valve MO-2316 will remain open (Bases EOP-1 pages 56 - 63).

C. Incorrect: The purpose of Defeat 8 is to permit the use of the RCIC turbine in order to 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 36

depressurize the RPV under non-line break conditions. The RPV High Water Level, RPV Low Steam Line Pressure, and High Ambient/Differential Temperature RCIC isolation signals are blocked with this Defeat. Defeat 8 is authorized for an alternate depressurization system utilized in Emergency Depressurization.

D. Incorrect: The RCIC Equipment room ambient temperature high isolation signal would not be present until 175F. 150F corresponds to the Suppression Pool Ambient Air temperature high setpoint (ARP 1C04C A-7).

Technical Reference(s): Bases-EOP 1, Revision 16 EOP 1, Revision 18 EOP Defeat 2, Rev. 3 ARP 1C04C, Revision 44 Proposed References to be provided to applicants during examination: N Learning Objective: 6.12.01 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments: Comments: SRO-only question justification is the link to 10CFR55.43(b)(5)

Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question covers Knowledge of diagnostic steps and decision points in the EOPs that involve transitions to event specific sub-procedures or emergency contingency procedures.

All control rods are fully inserted

HPCI suction swapped to the Torus

HPCI and RCIC tripped on high RPV level

RPV level is 195 inches and lowering

RPV pressure is 1060 psig and slowly rising

Drywell pressure is 1.5 psig Per the Emergency Operating Procedures, which EOP Defeat is needed to allow RCIC to be 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 37

used for control of RPV pressure?

A. Defeat 1, RCIC Low RPV Pressure Isolation and 211 inches Defeat B. Defeat 2, HPCI High Torus Water Level Transfer Defeat C. Defeat 8, RCIC Steam Line Isolation Defeat D. Defeat 18, HPCI/RCIC Area High Temp Isolation Defeat Answer: B 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 38

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A#: Generic K/A 2.1.32: Ability to explain and apply system limits and precautions.

Associated topic: 234000 (Fuel Handling Equipment).

Proposed Question: SRO Question # 91 Given the following:

The plant is in a refueling outage

The Fuel Handling Supervisor is checking prerequisites to commence fuel movement in accordance with IPOI-8, Outage and Refueling Operations, and RFP-403, Performance of Fuel Handling Activities

During the preceding plant shutdown, all rods were inserted on May 1, 2015 at time 0000

The outage schedule plans on having the core completely moved to the Fuel Pool by May 6, 2015 at time 0500

Fuel Pool water level is currently 36 feet, 11 inches

It is CURRENTLY May 3, 2015 at 0700 Which, if any, of the current conditions are not acceptable for fuel movement operations?

A. Conditions of time since shutdown, core movement rate, and Fuel Pool level are acceptable B. The fuel pool water level is too low to meet surveillance requirements C. The time since shutdown is too short to allow fuel movement at this time D. The scheduled rate of fuel movement is greater than the allowed rate Proposed Answer: C Explanation (Optional):

A. Incorrect: the time since the reactor has been shut down (defined as all rods full in) is required to be greater than 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months from the start of core alterations (RFP-403 page 6, TS Bases 3.9.6). Based upon the conditions in the stem, the current time since all rods were inserted is only 55 hours6.365741e-4 days 0.0153 hours 9.093915e-5 weeks 2.09275e-5 months .

B. Incorrect: whenever irradiated fuel is moved in the Spent Fuel Pool, pool level shall be maintained above 36 feet (RFP-403 page 6, TS 3.7.8). The level provided in the stem is however below the Fuel Pool low level setpoint of 37 feet, 1 inch (SD-435 pages 13-14).

C. Correct: the time since the reactor has been shut down (defined as all rods full in) is 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 39

required to be greater than 60 hours6.944444e-4 days 0.0167 hours 9.920635e-5 weeks 2.283e-5 months from the start of core alterations (RFP-403 page 6, TS Bases 3.9.6). Based upon the conditions in the stem, the current time since all rods were inserted is only 55 hours6.365741e-4 days 0.0153 hours 9.093915e-5 weeks 2.09275e-5 months .

D. Incorrect: RFP-403 states that the rate of discharge to the Spent Fuel Pool shall not exceed a rate that would result in the entire core being discharged within 121.33 hours3.819444e-4 days 0.00917 hours 5.456349e-5 weeks 1.25565e-5 months after shutdown. The rate of transfer during a core shuffle is not restricted because the entire core decay heat load will not be deposited into the Spent Fuel Pool (RFP-403 page 6). Based upon the conditions in the stem, fuel movement is scheduled to finish at the 125 hour0.00145 days 0.0347 hours 2.066799e-4 weeks 4.75625e-5 months point, which is acceptable.

Technical Reference(s): IPOI-8, Revision 82 TS 3.9.6, Amend. 280 TS Bases 3.9.6 RFP-403, Revision 54 SD-435, Revision 8 TS 3.7.8, Amend. 280 TS Bases 3.7.8 Proposed References to be provided to applicants during examination: None Learning Objective: 1.04.01.17 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments:

Comments: SRO-only question justification is the link to 10CFR55.43(b)(7) Fuel handling facilities and procedures.

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question covers Assessment of surveillance requirements for the refueling mode.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 40

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # A2.16 Importance Rating Ability to (a) predict the impacts of the following on the REACTOR/TURBINE PRESSURE REGULATING SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Low turbine inlet pressure (loss of pressure signal) 55.43(b)5 Proposed Question: SRO Question # 92 With the plant operating at 100% power, the pressure sensing line to Steam Throttle Pressure A transmitter ruptures resulting in 0 psig input to Pressure Regulator Channel A.

Which of the following will result from this failure, and what actions are required?

A. Reactor Pressure RISES 5 psig. It is required to enter AOP-262, Loss of Reactor Pressure Control and verify Core Thermal Limits B. Reactor Pressure LOWERS 5 psig. It is required to enter AOP-262, Loss of Reactor Pressure Control and verify Core Thermal Limits C. Reactor Pressure RISES, resulting in a Reactor Scram. It is required to enter EOP-1, RPV Control and verify all that control rods are inserted D. Reactor Pressure LOWERS, resulting in a Group 1 Isolation and a Reactor Scram. It is required to enter EOP-1, RPV Control and verify that all control rods are inserted Proposed Answer: A Explanation (Optional):

A. Correct: Reactor Pressure will RISE 5 psig. It is required to enter AOP-262 and verify Core Thermal Limits. With 0 psig input, a summer output goes to a large negative value, the HVG swaps over to the B regulator, which controls 5 psig lower due to the bias signal.

B. Incorrect: Identifies a misconception that bias channel results in 5 psig decrease.

C. Incorrect: This would be true if both A and B sensing taps ruptured.

D. Incorrect: This would be true if either Pressure Regulator output failed HIGH.

Technical Reference(s): AOP-262, Revision 7 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 41

Proposed References to be provided to applicants during examination: N Learning Objective: 52.01.01.02 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2007 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments: BANK - 2007 NRC Exam (SRO Question 92) 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 42

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A#: Generic K/A 2.2.42: Ability to recognize system parameters that are entry-level conditions for Technical Specifications. Associated topic: 271000 (Offgas).

Proposed Question: SRO Question # 93 Given the following:

The plant is operating at 30% power with a shutdown in progress

The shutdown is being conducted to support a Drywell entry to find the cause of increased leakage

Operators are performing an air purge (de-inerting) of the containment

Both Offgas Stack Radiation Monitors, RM-4116A and B, are declared inoperable due to a failed surveillance test procedure

KAMAN 9 and 10, Offgas Stack KAMAN monitors, remain in-service and operable Which of the following is true regarding the operators' ability to de-inert under these conditions?

De-inerting may A. NOT continue because containment venting in this situation would be an unmonitored release B. continue because the Offgas KAMANs being operable satisfy ODAM and Technical Specification requirements for a release C. NOT continue because a Group 3 isolation caused by RM-4116A and B inoperability would NOT allow containment venting D. continue as long as appropriate administrative controls are being maintained on the containment vent and purge valves while they are open Proposed Answer: D Explanation (Optional):

A. Incorrect: the ARP states RM-4116A&B are required in modes 1, 2, and 3 (during venting and purging of the Primary Containment. However T.S. Section 3.3.6.1.L.2 permits the use of alternate instrumentation.

B. Incorrect: Offgas KAMANs do not satisfy TS 3.3.6.1. They are part of TRM 3.3.3 instrumentation.

C. Incorrect: Offgas Stack Radiation Monitors, RM-4116A&B becoming inoperable do NOT cause a Group 3 isolation. Offgas Vent Pipe Radiation Monitors, RM-4116A&B Hi HI will 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 43

cause a Group 3 isolation but the downscale/inoperative does NOT.

D. Correct: RM-4116A&B are required in modes 1, 2, and 3 (during venting and purging of the Primary Containment. However T.S. Section 3.3.6.1.L.2 permits establishing administrative control of the primary containment vent and purge valves using continuous monitoring of alternate instrumentation.

Technical Reference(s): ARP 1C03A, Revision 53 DAEC TS 3.3.6.1, Primary Containment Isolation Instrumentation, Amendment 223 DAEC Technical Specification 3.3.6.1 Proposed References to be provided to applicants during examination:

(including Table 3.3.6.1-1)

Learning Objective: 1.02.01.01 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2011 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments: Comments: SRO-only question justification is the link to 10CFR55.43(b)(2)

Facility operating limitations in the Technical Specifications and their bases.

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question covers Application of Required Actions (Section 3) and Surveillance Requirements (Section 4) in accordance with rules of application requirements (Section 1).

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 44

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A#: Generic K/A 2.1.36: Knowledge of procedures and limitations involved in core alterations.

Proposed Question: SRO Question # 94 Given the following:

The plant is in a refueling outage

Preparations are being made to move fuel from the fuel pool back to the reactor vessel

The Fuel Handling Supervisor is reviewing prerequisites for core reload in accordance with RFP-403, Performance of Fuel Handling Activities

It has been identified that a change must be made to the approved Fuel Moving Plan (FMP)

Calculations have identified that the change WILL affect Shutdown Margin Based upon the conditions above, which of the following describes how the required change to the FMP will be made?

A. Made on the current FMP with Reactor Engineer approval B. Made on the current FMP with Fuel Handling Supervisor approval C. Made on the current FMP with Shift Manager approval D. Made only by fully revising the FMP Proposed Answer: D Explanation (Optional):

A. Incorrect: Minor pen & ink changes to the FMP may be made by Reactor Engineering with concurrence from the Fuel Handling Supervisor, Reactor Engineer, and the Shift Manager. Minor pen and ink changes are defined as any change that does not affect the Shutdown Margin as determined by the Reactor Engineering group (RFP-403 page 13).

B. Incorrect: Minor pen & ink changes to the FMP may be made by Reactor Engineering with concurrence from the Fuel Handling Supervisor, Reactor Engineer, and the Shift Manager. Minor pen and ink changes are defined as any change that does not affect the Shutdown Margin as determined by the Reactor Engineering group (RFP-403 page 13).

C. Incorrect: Minor pen & ink changes to the FMP may be made by Reactor Engineering with concurrence from the Fuel Handling Supervisor, Reactor Engineer, and the Shift Manager. Minor pen and ink changes are defined as any change that does not affect 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 45

the Shutdown Margin as determined by the Reactor Engineering group.

D. Correct: Any changes to the FMP that have the potential to affect the Shutdown Margin (SDM) as calculated by the Reactor Engineering Department will require a full revision to the FMP per REDP3, Core Alteration, and REDI 003, Creation of an Item Control Area (ICA) Transfer Report (RFP-403 page 13).

Technical Reference(s): RFP-403, Revision 54 Proposed References to be provided to applicants during examination: None Learning Objective: 1.04.01.06 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments: Comments: SRO-only question justification is the link to 10CFR55.43(b)(6)

Procedures and limitations involved in initial core loading, alterations in core configuration, control rod programming, and determination of various internal and external effects on core reactivity.

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question covers Administrative requirements associated with refueling activities, such as approvals required to amend core loading sheets or administrative controls of potential dilution paths and/or activities.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 46

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.1.20 Importance Rating Ability to interpret and execute procedure steps. CFR: 41.10 / 43.5 Proposed Question: SRO Question # 95 Given the following conditions:

The plant was operating at 100% reactor power

The station has suffered a station blackout

EOP-1, RPV Control, was entered

HPCI and RCIC are unavailable for injection

RPV water level is +20 inches and lowering at 1 inch per minute

ITC Midwest has informed the Control Room that offsite power will be restored within the next 30 minutes Which of the following actions is required at this time?

The SRO will direct an Emergency Depressurization .

A. IMMEDIATELY due to a lack of injection sources B. when RPV level lowers to +15 inches but before -25 inches C. when RPV level lowers to +15 inches and stop the depressurization prior to reaching 150 psig D. when RPV level lowers to -25 inches and stop the depressurization prior to reaching 150 psig Proposed Answer: B Explanation (Optional):

A. Incorrect: EOP 1 directs that an ED should be delayed to allow injection systems to be aligned.

B. Correct: If it is believed that available injection systems may not be capable of restoring and maintaining RPV water level above the Minimum Steam Cooling RPV Water Level (MSCRWL) following RPV depressurization, the blowdown should be delayed as long as possible.

C. Incorrect: EOP ED only requires that halting the ED to maintain steam driven high pressure injection system available. All steam driven injection systems are unavailable from conditions provided in the STEM.

D. Incorrect: EOP ED only requires that halting the ED to maintain steam driven high pressure injection system available. All steam driven injection systems are unavailable 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 47

from conditions provided in the STEM.

Technical Reference(s): EOP 1, Rev. 16 Proposed References to be provided to applicants during examination: N Learning Objective: 6.74.15.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments:

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 48

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A#: Generic K/A 2.2.44: Ability to interpret control room indications to verify the status and operation of a system, and understand how operator actions and directives affect plant and system conditions.

Proposed Question: SRO Question # 96 Given the following:

The plant was operating at 100% when a fuel leak resulted in high offgas and main steam line radiation levels

AOP 672.2, Offgas Radiation/Reactor Coolant High Activity, has been entered and a plant shutdown is being performed to comply with Technical Specifications

A spurious main turbine trip subsequently occurred and the plant automatically scrammed Current plant conditions are as follows:

ALL control rods are fully inserted

Reactor level lowered to 160" following the scram and is now stable at 184"

Reactor Pressure is 920 psig with the Turbine Bypass Valves in service

Offgas is in service, maintaining 2 inches Hg backpressure

1C05B(C-2), Main Steam Line HI HI RAD / INOP Trip, continues to alarm

EOP 1, RPV Control, has been entered on low RPV water level Based upon these conditions, which one of the following sets of actions is required AND will MINIMIZE release of radioactivity to the environment?

A. Enter EOP 4, Rad Release Control. Rapidly cooldown at GREATER THAN 100ºF/hr by depressurizing to the Main Condenser to allow the Offgas treatment process to limit radioactivity releases B. Enter EOP 4, Rad Release Control. Rapidly cooldown at GREATER THAN 100ºF/hr by depressurizing to the Torus to allow the Containment to limit radioactivity release and allow the Main Condenser to be used to control MSIV Leakage C. No additional EOP entries are required at this time. Cooldown at LESS THAN 100ºF/hr by depressurizing to the Main Condenser to allow the Offgas treatment process to limit radioactivity releases D. No additional EOP entries are required at this time. Cooldown at LESS THAN 100ºF/hr by depressurizing to the Torus to allow the Containment to limit radioactivity release and allow the Main Condenser to be used to control MSIV Leakage 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 49

Proposed Answer: D Explanation (Optional):

A. Incorrect: action would be correct if Emergency Depressurization were anticipated during EOP execution. No reasons are provided in stem for ED.

B. Incorrect: action would be correct if Emergency Depressurization were required and if EOP-4 Radioactivity Release Control, were entered. No entry conditions for these are given in stem.

C. Incorrect: action would be correct for a normal shutdown without High RCS Activity concerns.

D. Correct: AOP 672.2, Off Gas Radiation, Reactor Coolant High Activity specifies closing the MSIVs and MSL Drains, depressurizing to the Torus. Main Steam and Main Condenser will be aligned to limit MSIV Leakage. No requirement has been given to Anticipate Emergency Depressurization, so normal cooldown limits are in effect. EOP -

1 entry required on low RPV level, IPOI 5 entry not required because the scram already occurred (EOP 1 Decision Step RC-2) No other EOP entries exist.

Technical Reference(s): AOP-672.2, Revision 37 EOP-1, Revision 18 Proposed References to be provided to applicants during examination: None Learning Objective: 5.21.08.01 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments: Comments: SRO-only question justification is the link to 10CFR55.43(b)(4)

Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 50

Specifically, in reference to the Clarification Guidance for SRO-only Questions, this question covers Analysis and interpretation of radiation and activity readings as they pertain to selection of administrative, normal, abnormal, and emergency procedures.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 51

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A #

Importance Rating K/A#: Generic K/A 2.2.38: Knowledge of conditions and limitations in the facility license.

Proposed Question: SRO Question # 97 Given the following:

The plant is in MODE 2

Reactor Protection System (RPS) power is in its normal lineup The Instrument and Control (I&C) Supervisor reports that the Undervoltage trip for RPS Alternate Source Electrical Protection Assemblies (EPA) Power Supply breakers (C1 and C2) were set incorrectly and are outside their limit for OPERABILITY.

What action must be taken?

A. Be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B. Remove associated inservice power supply(s) from service within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months C. Remove associated inservice power supply(s) from service within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months D. Enter the alternate EPA breakers into the Technical Specification Inoperable Equipment Log Proposed Answer: D Explanation (Optional):

A. Incorrect - This would be required if the EPAs were in service when it was discovered and you exceeded the allowable out of service time.

B. Incorrect - This would be required if the EPAs were in service when it was discovered.

C. Incorrect - This would be required if the EPAs were in service when it was discovered.

D. Correct - Since the EPA breakers are not in-service the MODE of applicability for this technical specification is not met. The action would be to enter the components in the Technical Specification Inoperable Equipment Log.

Technical Reference(s): TS 3.3.8.2, Amend. 223 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 52

Proposed References to be provided to applicants during examination: TS 3.3.8.2 Learning Objective: 1.02.01.01 (As available)

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X Question History: Last NRC Exam:

Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments: Comments:

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 53

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.3.5 Importance Rating Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personal monitoring equipment, etc. (CFR 41.11/ 41.12/ 43.4)

Proposed Question: SRO Question # 98 The plant was at full power during day shift. While lowering a crate of highly radioactive material from the 5th floor, the sling broke, and the contents of the crate spilled out on the ground floor of the Reactor Building. No one was injured but the Railroad Access ARM is in alarm and reading 1000 times greater than normal

The OSM has declared EAL RU2.2, Any UNPLANNED VALID ARM reading GREATER THAN 1000 times normal The following has been completed in accordance with the NOTE 4, Plant Assembly Notification Form:

The outside speakers are turned on

The Evacuation Alarm was sounded

A plant page announcement has been made for all personnel to evacuate the Reactor Building

The Evacuation alarm and Plant Page announcement has been repeated Which of the following statements is correct in regard to the OSMs compliance with the Emergency Plan?

A. ALL OSM actions have complied with the NOTE 4 requirements B. The entire plant must be evacuated when the Evacuation Alarm is used for an EAL declaration C. The Evacuation Alarm is only used for EAL declarations of ALERT or greater, and may not be used for a Notification of Unusual Event D. The OSM should have declared a Notification of Unusual Event HU-5 (Other Conditions Exist Which in the Judgment of the Emergency Director Warrant Declaration of a NOUE), based on judgment Proposed Answer: B Explanation (Optional):

A. Incorrect: Per EPIP 1.3,in an EAL condition, the entire plant must be evacuated for 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 54

accountability purposes.

B. Correct: Per EPIP 1.3 and EPIP FORM NOTE 4, the wrong announcement for plant assembly was made.

C. Incorrect: Evacuation alarm must be sounded for ALERT or greater, but may also be used for general evacuation or NUEs.

D. Incorrect: On-site radiation levels have met RU2.2, HU-5 would be incorrect for the given conditions.

Technical Reference(s): EPIP 1.3, Revision 19.

Proposed References to be provided to applicants during examination: N Learning Objective: 3.01.03.02 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2001 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments: BANK - 2001 NRC Exam 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 55

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2.4.16 Importance Rating 4.4 Knowledge of EOP implementation hierarchy and coordination with other support procedures or guidelines such as, operating procedures, abnormal operating procedures, and severe accident management guidelines. (CFR 41.10/43.5)

Proposed Question: SRO Question # 99 Given the following:

The EOPs have been entered and plant conditions have degraded such that SAG entry is required

The TSC is NOT ready to assume control Which of the following is correct?

The operating crew should A. continue implementing the current EOP actions until the TSC is ready to transition to the SAGs.

B. exit the EOP which directs the entry into the SAGs and continue to implement all other EOPs which are entered.

C. exit the EOP leg that is directing the SAG entry and continue to implement all other EOPs legs in effect.

D. enter the SAG that is directed and when the TSC is ready, turnover all actions which were directed from the SAGs entered.

Proposed Answer: A Explanation (Optional):

A. Correct: Until the TSC is ready the operating crew is directed to continue to use the EOP strategies to combat the event.

B. Incorrect: Exiting the EOPs is correct. However, the TSC must be ready to take control and ALL EOPs are exited at that time.

C. Incorrect: Exiting the EOPs is correct. However, the TSC must be ready to take control and ALL EOPs are exited at that time.

D. Incorrect: Entering the SAGs would be correct if the TSC is ready. The crews do not enter the SAGs without the TSC being ready.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 56

Bases-EOP Flow chart use and Technical Reference(s):

logic, Rev 4 page 43 Proposed References to be provided to applicants during examination: N Learning Objective: 95.74.16.01/95.74.16.02 (As available)

Explain the transition process from EOPs to SAGs/Explain the concept of default actions as it pertains to the actions to take while still in EOPs and waiting to make the transition to SAGs Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2002 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 Comments: BANK - DAEC 2002 NRC Exam SRO Basis: CFR: 43.5 - Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 57

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # 2.4.35 Importance Rating Knowledge of local auxiliary operator tasks during an emergency and the resultant operational effects. CFR: 41.10 / 43.5 / 45.13 Proposed Question: SRO Question # 100 Given the following:

The plant was operating at full power

The control room had to be evacuated due to a fire

All required control room actions were completed prior to the evacuation Which one of the following describes a task that must be completed IAW AOP 915, Shutdown Outside the Control Room?

A. Transfer control to Remote Shutdown Panel 1C388 ONLY within 20 minutes B. Transfer control to Remote Shutdown Panels 1C388 AND 1C389 within 20 minutes C. Establishing additional ventilation in the 1A4 switchgear room ONLY within 30 minutes D. Establishing additional ventilation in the 1A3 switchgear rooms ONLY within 30 minutes Proposed Answer: B Explanation (Optional):

A. Incorrect: IAW AOP 915, transfer of panel 1C389 is also required for SRV control.

B. Correct: IAW AOP 915, per Caution on Page 6, For Control Room evacuation as the result of a fire, transfer of control at panels 1C388, 1C389, 1C390, 1C391, 1C392 is required to be completed within 20 minutes.

C. Incorrect: The requirement has no time constraints.

D. Incorrect: The requirement has no time constraints.

Technical Reference(s): AOP-915, Revision 53 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 58

Proposed References to be provided to applicants during examination: N Learning Objective: 5.28.01.02 (As available)

Question Source: Bank # X Modified Bank # (Note changes or attach parent)

New Question History: Last NRC Exam: 2009 Duane Arnold Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 Comments: BANK (DAEC 2009 NRC Exam) 50007/60006 Rev. 0 PDA 15-1 RO/SRO NRC Written exam 60006_PDA 15-1 RO-SRO NRC Written_xm Page 59

WRITTEN / ORAL EXAMINATION KEY Page 1 COVER SHEET Examination Number/Title: PDA 15-1 RO NRC Written Exam Training Program: Licensed Operator Initial Training Course/Lesson Plan Number(s): 50007 Reactor Operator Total Points Possible: 75 PASS CRITERIA: 80%. Exam Time: 360 minutes Yes No Yes No This is an alternate examination; verified This is a remediation exam; verified at at least 30% of the questions are least 90% of the questions are different different from other forms/versions of this from the failed exam. For LOIT remedial exam (e.g., Forms A, B, C; continuing exams, verified 95% difference. For training exam versions for consecutive LOCT annual operating and biennial weeks). For LOCT weekly exams during comprehensive remedial exams, verified a segment, verified > 50% difference. no repeat questions.

This is an initial training examination; This is a randomly generated electronic verified at least 30% of the questions are exam printout; verified the exam bank has different from same exam administered 3 questions per objective if one test item to the previous class. on exam for the objective. If 2 or more test items on exam for an objective, then 6 questions are in bank.

Exam development and review guidelines: Key should contain the following:

o TR-AA-230-1003, SAT Development Learning Objective Number Test Item o Question or Statement o All possible answers o Correct Answer Indicated o Point Value o References (if applicable)

NOTE: NRC exams may require additional information. Refer to fleet and site specific procedures.

EXAMINATION REVIEW AND APPROVAL:

Developed by: Date:

Instructional Review of Written Exam (Qualified Instructor): Date:

Technical Review (SME): Date:

Approved by Training Supervisor: Date:

Approved by Training Program Owner (or line designee): Date:

Indicate in the following table if any changes are made to the exam after approval:

AR/TWR# PREPARER DATE

DESCRIPTION OF CHANGE REASON FOR CHANGE (if applicable) SUPERVISOR DATE Filename: 50007_PDA 15-1 RO NRC Written_xm TR-AA-230-1003-F13 Revision 1