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Dua 2017 Ile - Proposed Exam SRO Written Examination
	ML17214A050
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Site:	Duane Arnold
Issue date:	06/14/2017
From:	Bielby M; Operations Branch III
To:	; NextEra Energy Duane Arnold
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ML16308A458	List: ML17214A027; ML17214A028; ML17214A034; ML17214A036; ML17214A038; ML17214A050;
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WRITTEN / ORAL / ONLINE EXAMINATION KEY Page 1 of 1 COVER SHEET Examination Number/Title: PDA 17-1 NRC Exam SRO, Rev. 0 Training Program: Operations Course/Lesson Plan Number(s): 60006 Total Points Possible: 25 PASS CRITERIA: 70% Exam Time: 120 Yes No Yes No This is an alternate examination; verified This is a remediation exam. Verified the at least 30% of the questions are questions are different from the failed different from other forms/versions of this exam by at least the following criteria exam (e.g., Forms A, B, C; continuing listed below:

training exam versions for consecutive

70% for Maintenance/Technical weeks). For LOCT weekly exams during a segment, verified > 50% difference.

90% for Operations training programs This is an initial training examination; This is a LOCT annual operating exam or verified at least 30% of the questions are biennial comprehensive remedial exam, different from previous administration of verified the questions are 100% different the same exam. from the failed exam.

This is a non-randomly generated exam This is a randomly generated exam from from an electronic exam bank, printed an electronic exam bank, printed out or out or administered online. Verified at administered online. Verified the exam least 30% of the questions are different bank has 3 questions per objective if one from other forms/versions of this exam test item on exam for the objective. If 2 or (e.g., Forms A, B, C; continuing training more test items on exam for an objective, exam versions for consecutive weeks). then 6 questions are in bank.

For LOCT weekly exams during a segment, verified > 50% difference.

NOTE: Key should contain the following:

Learning Objective Number

See TR-AA-230-1003, SAT Test Item Development, for exam development o Question or Statement and review guidelines.

o All possible answers

NRC exams may require additional o Correct Answer Indicated information. Refer to fleet and site o Point Value specific procedures. o References (if applicable)

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 17-1 NRC Exam SRO_xm_60Day.docx TR-AA-230-1003-F13, Revision 6

WRITTEN / ORAL EXAMINATION COVER SHEET Page 1 Trainee Name:

Employee Number: Site: PDA Examination Number/Title: PDA 17-1 NRC Exam SRO, Rev. 0 Training Program: Operations Course/Lesson Plan Number(s): 60006 Total Points Possible: 25 PASS CRITERIA: 70% Grade: /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 120 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:

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 1

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295001 AA2.04 Importance Rating 3.1 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF FORCED CORE FLOW CIRCULATION: Individual jet pump flows: Not-BWR-1&2 Proposed Question: SRO Question 1 (76)

The plant was operating at full reactor power when an event occurred. The Operator at the Controls reported that Total Core Flow lowered from 49.5 Mlbm/hr to 44 Mlbm/hr.

The following indications were observed, after the event, at 1C38:

What action is required?

A. Establish single loop operations within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 2

B. Be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

C. TRIP the A Recirc pump in 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months D. Be in MODE 4 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months Proposed Answer: B Explanation (Optional):

A. Incorrect - IF the applicant assumes the indications are due to a pump trip, single loop operations are required to be established within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months per LCO 3.4.1.

B. Correct - These indications are consistent with Jet Pump failure. This condition requires the plant to be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months per LCO 3.4.2.

C. Incorrect - If the applicant believes a speed mismatch due to a recirc pump runback has occurred, the action would be to trip the running pump within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months per LCO 3.4.1 D. Incorrect - The applicant may perceive an unanalyzed condition based upon the indications, the applicant may elect to enter LCO 3.0.3 which requires MODE 4 in 37 hours4.282407e-4 days 0.0103 hours 6.117725e-5 weeks 1.40785e-5 months .

AOP 255.2 Power/Reactivity Technical Reference(s): Abnormal Change rev. 43 Bases (Attach if not previously provided)

LCO 3.4.2 Safety Analysis rev. 2 Proposed References to be provided to applicants during examination: L3.4.1, L3.4.2 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 5 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 3

Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 4

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295016 AA2.02 Importance Rating 4.3 Ability to determine and/or interpret the following as they apply to CONTROL ROOM ABANDONMENT: Reactor Water Level Proposed Question: SRO Question 2 (77)

The plant was operating at full reactor power when a fire in the cable spreading room has resulted in heavy smoke in the control room.

All efforts to extinguish the fire have been unsuccessful

An immediate evacuation of the control room is required Transfer of control to the remote shutdown panel must be completed within 20 minutes to A. close a spuriously opened SRV B. raise RPV water level for natural circulation C. depressurize the RPV to its lowest energy state D. prevent spurious operation of the ECCS Low Pressure Injection Pumps Proposed Answer: A Explanation (Optional):

A. Correct - Control of Time Critical Tasks states that transfer of 1C388 must be completed within 20 minutes to close a spuriously opened SRV to control RPV inventory in this condition...

B. Incorrect - Plausible since it is a mitigating action for loss of core flow and to promulgate flow through the core. In the given circumstance however, these conditions are not present. In addition this is not the basis for the time critical operator action.

C. Incorrect - while plausible since this is a mitigating action, it is not the basis for the time critical operator action.

D. Incorrect - while plausible since this is a mitigating action, it is not the basis for the time critical operator action.

ACP 103.10 Control of Time Technical Reference(s): (Attach if not previously provided)

Critical Tasks, Rev. 11 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 5

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 10 CFR Part 55 Content: 55.41 55.43 1 Comments: Conditions and limitations in the facility license.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 6

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295019 AA2.02 Importance Rating 3.7 Ability to determine and/or interpret the following as they apply to PARTIAL OR COMPLETE LOSS OF INSTRUMENT AIR: Status of safety-related instrument air system loads Proposed Question: SRO Question 3 (78)

The plant is operating at full reactor power with 1K4, CB/SBGTS Instrument Air Compressor, INOPERABLE.

1K4 Day 2 of maintenance for planned repairs

At 1600 1A311, 1A3 Bus Feeder Breaker from SBDG 1G-31 indications have been lost

At 1630 Electricians determine that the control power fuses for 1A311 have been blown

Repairs cannot be completed for 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months What action is the CRS first required to comply with?

A. Declare 1K3 INOPERABLE immediately B. Declare 1K3 INOPERABLE by 2000 C. Restore 1K4 to OPERABLE within 7 days D. Restore A SBDG 1G-31 to OPERABLE by 1600 the following day.

Proposed Answer: B Explanation (Optional):

A. Incorrect - 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is allotted when the required redundant feature is B. Correct - 1K4 is INOPERABLE. With the A SBDG INOPERABLE, 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is allotted before the required redundant feature (1K3) must be declared INOPERABLE.

C. Incorrect - Although still in this specification, there are other actions that are required to be first taken.

D. Incorrect - This may be true as well however there are other actions that are required to be taken first.

LCO 3.8.1 AC Sources -

Operating rev. 272, LCO 3.7.9 Technical Reference(s): (Attach if not previously provided)

CB/SBGT Instrument Air System rev. 229 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 7

Proposed References to be provided to applicants during examination: L3.7.9, L3.8.1 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 2 Comments: Facility operating limitations in the technical specifications and their bases.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 8

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295023 2.2.22 Importance Rating 4.7 Knowledge of limiting conditions for operations and safety limits: Refueling Accidents Proposed Question: SRO Question 4 (79)

The plant is in MODE 5 with the fuel pool gates installed for a RFO with the following conditions:

Fuel is being moved in the Spent Fuel Pool Then, annunciator FUEL POOL COOLING PANEL 1C-65/1C-66 TROUBLE 1C04B (D-2) alarms. The following plant conditions exist:

The cause of the alarm is Skimmer Surge Tank Low Level

Spent Fuel Pool level is 35 feet and slowly lowering

Refuel Floor ARMs have increased by 2 mrem / hr and not alarming (1) What actions are required?

AND (2) What, if any, EAL is required to be declared?

A. (1) Movement of irradiated fuel assemblies in the spent fuel storage pool may continue indefinitely provided contingencies in place.

(2) An Alert must be declared.

B. (1) Movement of irradiated fuel assemblies in the spent fuel storage pool must be suspended immediately.

(2) An Unusual Event must be declared.

C. (1) Movement of irradiated fuel assemblies in the spent fuel storage pool must be suspended immediately.

(2) An Alert must be declared.

D. (1) Movement of irradiated fuel assemblies in the spent fuel storage pool may continue indefinitely provided contingencies in place.

(2) An Unusual Event must be declared.

Proposed Answer: B Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 9

Explanation (Optional):

A. Incorrect -The LCO has been exceeded and UE is required IAW EAL-02, RU2.1, Unplanned valid Refuel Floor ARM reading increase with an uncontrolled loss of reactor cavity, fuel pool, or fuel transfer canal water level with all irradiated fuel assemblies remaining covered by water as indicated by any of the following:

Report to control room

Valid fuel pool level indication (LI-3413) LESS THAN 36 feet and lowering

Valid WR GEMAC Floodup indication (LI-4541) coming on scale B. Correct - IAW ARP 1C04B A-4 Section 3.7 & 3.8. TS 3.7.8 LCO limit is 36 feet, action is required to immediately suspend fuel movement. An Unusual Event must be declared because fuel pool level indication (LI-3413) LESS THAN 36 feet and lowering.

C. Incorrect - A UE must be declared D. Incorrect - Movement of irradiated fuel must be suspended immediately.

TS 3.7.8 Spent Fuel Storage Pool water level rev. 224, EPIP Form Technical Reference(s): (Attach if not previously provided)

EAL-02 EAL Matrix - Cold MODE, Rev 9 EPIP Form EAL-01 Proposed References to be provided to applicants during examination:

EPIP Form EAL-02 Learning Objective: (As available)

Question Source: Bank #

Modified Bank # X (Note changes or attach parent)

New Question History: Last NRC Exam: 2011 NRC Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 2 Comments: Facility operating limitations in the technical specifications and their bases.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 10

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295025 EA2.03 Importance Rating 4.1 Ability to determine and/or interpret the following as they apply to HIGH REACTOR PRESSURE: Suppression pool temperature Proposed Question: SRO Question 5 (80)

The following conditions were present following a Reactor SCRAM:

Reactor pressure is 910 psig and stable

Torus water level is 10.4 feet and stable

Torus water temperature is 1600F and rising slowly

Drywell pressure is 6 psig and rising slowly

Drywell air temperature is 2200F and rising slowly The CRS will____________________.

A. anticipate emergency depressurization B. emergency depressurize C. lower reactor pressure 200 psig D. vent the Drywell Proposed Answer: B Explanation (Optional):

A. Incorrect - The HCL curve has been exceeded. This requires emergency depressurization.

B. Correct - Since the HCL curve has been exceeded, emergency depressurization is required.

C. Incorrect - Since the HCL curve has been exceeded, emergency depressurization is required.

D. Incorrect - Group 3 isolation prevents venting the drywell. Venting containment is not yet required.

EOP 1 RPV Control rev.20, EOP 2 Technical Reference(s): Primary Containment Control rev. (Attach if not previously provided) 18 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 11

Proposed References to be provided to applicants during examination: EOP Graph 4 - HCL 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 5 Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 12

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295028 G2.4.8 Importance Rating 4.5 Knowledge of how abnormal operating procedures are used in conjunction with EOPs.: High Drywell Temperature Proposed Question: SRO Question 6 (81)

A loss of offsite power occurs and the following conditions are present:

1A3 and 1A4 4160VAC Essential Busses are being supplied by A and B SBDGs

Drywell Pressure is 6 psig rising slowly

Drywell air temperature is 2800F and rising slowly What if any action(s) must the CRS direct?

A. No additional actions are required B. Emergency Depressurize the RPV C. Initiate Drywell Sprays D. Restore power to 1B33, 1B43, 1B35, and 1B45 Proposed Answer: D Explanation (Optional):

A. Incorrect - to restore drywell cooling, power must be restored to Well Water and to drywell cooling fans.

B. Incorrect - This would be true if drywell air temperature exceeded 340F C. Incorrect - Outside the allowable Drywell spray initiation limit curve.

D. Correct - LOOP/LOCA Load shedding has occurred requiring restoration of power to 1B33,35,43, and 45. AOP 301 directs these actions.

AOP 301 Loss of essential electrical power, rev. 71 Technical Reference(s): (Attach if not previously provided)

Restoration of Power to essential busses. Page 34 step d(1)

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 13

EOP Graph 7 -

Proposed References to be provided to applicants during examination:

DWSIL 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 5 Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 14

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 1 K/A # 295037 G2.4.34 Importance Rating 4.1 Knowledge of RO tasks performed outside the main control room during an emergency and the resultant operational effects: SCRAM Condition Present and Reactor Power Above APRM Downscale or Unknown Proposed Question: SRO Question 7 (82)

An Electrical ATWS has occurred.

Both SBLC Pumps fail to run from the control room The CRS shall direct __________________ to inject Boron to the RPV.

A. OI 153 Standby Liquid Control System B. AIP 406 Injection with SBLC C. SEP 304 Boron Injection using RWCU D. OI 261 Reactor Water Cleanup System Proposed Answer: C Explanation (Optional):

A. Incorrect - The Operating instruction does not provide the procedural guidance for alternate injection of boron but does provide direction to inject boron to the RPV.

B. Incorrect - The AIP could inject Boron into the RPV however it requires the use of SBLC pumps which are not available C. Correct - per EOP ATWS step /Q-D. Incorrect - Although RWCU system is used to inject Boron to the RPV, the Operating instruction does not provide the procedural guidance for alternate injection of boron EOP ATWS rev. 23, SEP 304 Technical Reference(s): Boron Injection using RWCU rev. (Attach if not previously provided) 16 Proposed References to be provided to applicants during examination: N Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 15

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 5 Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 16

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295007 G2.1.20 Importance Rating 4.6 Ability to interpret and execute procedure steps: High Reactor Pressure Proposed Question: SRO Question 8 (83)

10:00 While performing a plant walkdown at full reactor power, the Balance of Plant (BOP) Operator reports that the B EHC Pressure Regulator is in service.

10:02 The At the Controls Operator reports that Reactor Pressure is 1030 psig.

10:17 Attempts to lower Reactor Pressure were unsuccessful.

(1) The CRS will direct _________________ to mitigate this condition?

AND (2) What is the first action that must be completed?

A. (1) AOP 693 Main Turbine/EHC Failures (2) Be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months B. (1) AOP 693, Main Turbine/EHC Failures (2) Reduce Thermal Power to <21.7% within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months C. (1) AOP 262, Loss of Reactor Pressure Control (2) Be in MODE 3 within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months D. (1) AOP 262, Loss of Reactor Pressure Control (2) Reduce Thermal Power to <21.7% within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months Proposed Answer: D Explanation (Optional):

A. Incorrect - AOP 693 does not provide the guidance for a failed pressure regulator. The first action required to be completed is thermal power to <21.7% within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in accordance with LCO 3.2.2 B. Incorrect - AOP 693 does not provide the guidance for a failed pressure regulator.

C. Incorrect - The first action required to be completed is thermal power to <21.7% within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months in accordance with LCO 3.2.2 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 17

D. Correct - AOP 262 provides the guidance for a failed pressure regulator. Step 6 directs lowering power <21.7% to comply with the MCPR spec 3.2.2.

AOP 262 Loss of Reactor Pressure Control, rev. 8 step 6, Technical Reference(s): LCO 3.2.2 MCPR rev. 244, LCO (Attach if not previously provided) 3.4.10 Reactor Steam Dome Pressure Rev. 224 L3.2.2 Proposed References to be provided to applicants during examination:

L3.4.10 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 2 Comments: Facility operating limitations in the technical specifications and their bases.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 18

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295010 G2.4.21 Importance Rating 4.6 Knowledge of the parameters and logic used to assess the status of safety functions, such as reactivity control, core cooling and heat removal, reactor coolant system integrity, containment conditions, radioactivity release control, etc. : High Drywell Pressure Proposed Question: SRO Question 9 (84)

Station Blackout conditions exist

The RPV has been Emergency Depressurized and Reactor Pressure is being maintained 150-350 psig

Torus Water level is 12.2 feet and rising slowly

Containment Pressure is 40 psig and rising slowly

All available support systems are operating as designed The CRS directs ______________________ to lower containment pressure to 10 psig.

A. SEP 307, Rapid Depressurization with Bypass Valves B. SEP 301.3, Torus Vent via Hard Pipe Vent C. SEP 301.2, Drywell Vent via SBGT D. SEP 301.1, Torus Vent via SBGT Proposed Answer: B Explanation (Optional):

A. Incorrect - This action would remove energy from the containment making it plausible.

B. Correct - With RPS lost, the other choices are not possible. The CRS would direct use of the Torus Hard Pipe Vent to lower containment pressure.

C. Incorrect - RPS is not available. Torus venting is preferable if available D. Incorrect - RPS is not available.

EOP 2 Primary Containment Control rev. 18, SEP 301.1 Torus Technical Reference(s): Vent via SBGT rev. 10, SEP 301.2 (Attach if not previously provided)

Drywell Vent via SBGT rev. 6, SEP 301.3 Torus Vent via Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 19

Hardpipe vent, rev. 10 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 5 Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 20

Examination Outline Cross-reference: Level RO SRO Tier # 1 Group # 2 K/A # 295014 AA2.02 Importance Rating 3.9 Ability to determine and/or interpret the following as they apply to INADVERTENT REACTIVITY ADDITION: Reactor period Proposed Question: SRO Question 10 (85)

During a reactor startup, control rod 26-27 is withdrawn one notch. Reactor period is observed to change from 150 seconds to 50 seconds.

Which one of the following describes the appropriate action to be taken?

The CRS should direct the RO to ____________________.

A. immediately SCRAM the reactor B. re-insert control rods as necessary to go to a subcritical hold point C. shutdown the reactor until a thorough assessment can be performed D. re-insert control rod 26-27 to obtain a stable period of greater than 60 seconds Proposed Answer: D Explanation (Optional):

A. Incorrect - Reactor Scram is not required at this time.

B. Incorrect - Inserting control rods to a subcritical hold point is only valid if the reactor remained subcritical C. Incorrect -.Reactor shutdown is not required for a double notched rod.

D. Correct - The CRS would direct the RO to reinsert control rods as necessary to maintain a manageable period.

Technical Reference(s): IPOI 2 Startup P&L 12 rev 154, (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 21

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 5 Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 22

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 206000 A2.02 Importance Rating 3.5 Ability to (a) predict the impacts of the following on the HIGH PRESSURE COOLANT INJECTION SYSTEM ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Valve closures:

BWR-2,3,4 [High Pressure Coolant Injection (HPCI) System]

Proposed Question: SRO Question 11 (86)

The plant is in MODE 2 and Reactor Pressure is 940 psig and stable

A through wall piping leak is identified in the HPCI steam supply line.

The leak has been isolated by closing MO2238 and MO2239 HPCI Steam Supply Isolation Valves.

Mode change to MODE 1 is __________________.

A. allowed by LCO 3.0.4a B. allowed by LCO 3.0.4b C. allowed by LCO 3.0.4c D. not allowed Proposed Answer: D Explanation (Optional):

A. HPCI is INOPERABLE when isolated. MODE change to MODE 1 is not allowed.

B. HPCI is INOPERABLE when isolated. MODE change to MODE 1 is not allowed.

C. HPCI is INOPERABLE when isolated. MODE change to MODE 1 is not allowed.

D. Correct: A throughwall leak in the HPCI Steam Supply Line LCO 3.7.3 Condition A applies. Isolate the piping to exit the applicability to the action statement. Isolating HPCI makes HPCI INOPERABLE, MODE change to MODE 1 is not allowed with LCO 3.5.1 Condition .

TRM 3.7.3 Structural Integrity rev.

8, LCO 3.5.1Emergency Core Technical Reference(s): (Attach if not previously provided)

Cooling Systems (ECCS) 259, LCO 3.0.4 LCO Applicability rev.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 23

266 Proposed References to be provided to applicants during examination: N 5.01.01.07, State when the HPCI System is required to be OPERABLE Learning Objective: by Technical Specifications and (As available) describe the bases of the HPCI System LCOs 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 2 Comments: Facility operating limitations in the technical specifications and their bases.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 24

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 209001 G2.2.37 Importance Rating 4.7 Ability to determine operability and/or availability of safety related equipment: Low Pressure Core Spray System Proposed Question: SRO Question 12 (87)

The plant is operating at full reactor power

RHR Pump B supply breaker is racked down and removed for maintenance on the breaker

During testing it is determined that the B Core Spray Pump could not develop more than 2650 gpm at 110 psig.

Which one of the following satisfies the most limiting action?

A. Enter LCO 3.0.3 immediately B. No additional actions are required C. Restore RHR Pump B to OPERABLE within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months D. Restore Core Spray Pump B to OPERABLE in 7 days Proposed Answer: C Explanation (Optional):

A. Incorrect - Since LCO 3.5.1 condition N is modified by a note that allows for condition C therefore performing this action is overly conservative and not in accordance with provisions allowed in the spec.

B. Incorrect - The data provided makes the B Core Spray Pump INOPERABLE therefore additional actions are required C. Correct - With 2 Low Pressure ECCS Pumps INOPERABLE, a 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months LCO is required to be entered. Restoration of one of the 2 pumps to OPERABLE is the most restrictive action for the given conditions.

D. Incorrect - This is a required action however, it is not the most limiting action required at this time.

LCO 3.5.1Emergency Core Technical Reference(s): Cooling Systems (ECCS) rev.259 (Attach if not previously provided)

SR3.5.1.4 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 25

Proposed References to be provided to applicants during examination: L3.5.1 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 2 Comments: Facility operating limitations in the technical specifications and their bases.

Had to redraw the KA. Unable to write to original. EM 3/22/2017 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 26

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 212000 A2.02 Importance Rating 3.9 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: RPS bus power supply failure Proposed Question: SRO Question 13 (88)

The plant was operating at full reactor power when an electrical transient occurred resulting in an automatic reactor scram.

The following are plant conditions:

All control rods are fully inserted

Reactor level has recovered to 178 inches and is slowly rising

Reactor Pressure is 1000 psig and rising slowly

The Main Turbine is tripped

The SBDGs are running unloaded with the Essential Buses on their normal supply

RPS is deenergized The CRS shall direct which of the following pressure control strategies?

A. Cool down the RPV with the Main Turbine Bypass Valves utilizing EHC pressure set.

Defeat isolation interlocks as necessary B. Place RWCU in operation using SEP 302.1, RWCU in Recirc mode, or using SEP 302.2, RWCU in Drain mode C. Place RCIC or HPCI in pressure control mode. Defeat isolation interlocks as necessary D. Install Defeat 17 Hi Condenser Backpressure Bypass and utilize MSL Drains MO-4423 and MO-4424 for pressure control Proposed Answer: C Explanation (Optional):

A. Incorrect - A group 1 isolation is present with the loss of RPS power making this pressure control mode unavailable B. Incorrect - RWCU is not available due to the loss of RPS power and the system being isolated Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 27

C. Correct - HPCI and RCIC are available in pressure control mode There is no 2 psig Drywell pressure signal making this a possible strategy D. Incorrect - RPS power is lost. Cannot override the isolation signal.

Technical Reference(s): EOP 1 RPV Control rev.20. (Attach if not previously provided)

Proposed References to be provided to applicants during examination: N 5.11.01.02, Evaluate plant conditions and control room indications and Learning Objective: (As available) determine necessary mitigative strategies (AOP 358)

Question Source: Bank # 50632 Modified Bank # (Note changes or attach parent)

New PDA 15-1 Workup Question History: Last NRC Exam:

Comp Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 28

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 217000 2.4.6 Importance Rating 4.7 Knowledge of EOP mitigation strategies: Reactor Core Isolation Cooling (RCIC) System Proposed Question: SRO Question 14 (89)

An event has occurred and the plant is in a station blackout condition. The following are plant conditions:

Reactor pressure is being maintained 150 to 350 psig

Reactor Core Isolation Cooling (RCIC) has been maximized for reactor vessel injection

HPCI is unavailable

Drywell air temperature will reach 340°F in 90 minutes

Torus water temperature is rising such that Heat Capacity Limit (HCL) will be exceeded in 45 minutes

Two unisolable steam leaks are raising local area temperatures and will exceed max safe in both areas in 55 minutes

Portable Diesel Driven Fire Pump will have reactor vessel injection capabilities in 70 minutes Given these plant conditions and times, when is the Control Room Supervisor first required to perform a full reactor vessel depressurization?

A full reactor depressurization will be required when ____________________.

A. alternate injections sources are available B. two area temperatures are greater than max safe C. the Heat Capacity Limit curve has been violated D. drywell air temperature reaches 340°F Proposed Answer: A Explanation (Optional):

A. Correct - Full depressurization is not allowed until alternate injections sources are available B. Incorrect - Continue full depressurization when two areas are greater than max safe C. Incorrect - Continue full depressurization when HCL has been exceeded Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 29

D. Incorrect - Continue full depressurization when 340F EOP Emergency RPV Technical Reference(s): (Attach if not previously provided)

Depressurization, rev. 11 EOP 3 Secondary Containment /

Rad Release Control, rev. 22 EOP 2 Primary Containment Control rev. 18 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 5 Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 30

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 1 K/A # 262001 A2.06 Importance Rating 2.9 Ability to (a) predict the impacts of the following on the A.C. ELECTRICAL DISTRIBUTION ;

and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Deenergizing a plant bus Proposed Question: SRO Question 15 (90)

Following a loss of Essential AC power, the crew has entered the LOSS OF BOTH ESSENTIAL 4160V BUSES section of AOP 301, LOSS OF ESSENTIAL ELECTRICAL POWER.

Which ONE of the following correctly describes:

(1) The impact on Station Fire Protection?

AND (2) The procedure which dictates the establishment of "compensatory" fire patrols?

A. (1) All Fire Detection Systems become NON-FUNCTIONAL (2) EN-AA-202-1004, FIRE PROTECTION SCREENING B. (1) Loss of Fire Barrier when fire doors are blocked open (2) EN-AA-202-1004, FIRE PROTECTION SCREENING C. (1) Loss of Fire Barrier when fire doors are blocked open (2) ACP 1412.4, IMPAIRMENTS TO FIRE PROTECTION SYSTEMS D. (1) All Fire Detection Systems become NON-FUNCTIONAL (2) ACP 1412.4, IMPAIRMENTS TO FIRE PROTECTION SYSTEMS Proposed Answer: C Explanation (Optional):

A. Incorrect - Some fire detection systems are supplied by their own batteries and would not be INOPERABLE and therefore would not require "compensatory" fire patrols.

B. Incorrect - EN-AA-202-1004, FIRE PROTECTION SCREENING. and the DAEC Fire Plan describe the overall fire protection program at the Duane Arnold Energy Center.

There is no specific guidance for impairments in EN-AA-202-1004, FIRE PROTECTION SCREENING..

C. Correct - IAW AOP-301, page 11, Consult with Security, establish natural/temporary ventilation by opening room/panel doors as follows: Battery Room doors, Nonessential Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 31

Switchgear Room doors, Control Room doors, and Essential Switchgear Room doors.

A caution in this step is "This step affects the integrity of some plant fire zones. FPIRs and fire watches may be required." ACP-1412.4, Impairments to Fire Protection Systems, provides the requirement for establishing "compensatory" fire patrols.

D. Incorrect - Some fire detection systems are supplied by their own batteries and would not be INOPERABLE and therefore would not require "compensatory" fire patrols.

AOP 301 Loss of Essential Technical Reference(s): (Attach if not previously provided)

Electrical Power, Rev. 71 ACP 1412.4, Rev.

Proposed References to be provided to applicants during examination: N 5.04.01.02, Evaluate plant conditions and control room indications and Learning Objective: (As available) determine the actions directed by AOP 301 Question Source: Bank # 50469 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 2 Comments: Facility operating limitations in the technical specifications and their bases.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 32

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 201006 G2.2.40 Importance Rating 4.7 Ability to apply Technical Specifications for a system: Rod Worth Minimizer System (Plant Specific)

Proposed Question: SRO Question 16 (91)

Reactor Power is at 8% with the Rod Worth Minimizer out of service. Operators inserting control rods inadvertently skipped several rod sequence control pages resulting in seven rods out of the allowable sequence.

Which one of the following is the maximum time to correct the condition?

A. 2 Hours B. 4 Hours C. 8 Hours D. 12 Hours Proposed Answer: C Explanation (Optional):

A. Incorrect - The applicant could correlate a violation of thermal limits which has a 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months completion time to correct the condition B. Incorrect - The applicant could consider LCO 3.1.3 condition D which has a 4 hour4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months completion time for 2 or more INOPERABLE control rods not within the BPWS C. Correct - LCO 3.1.6 condition A required action A.1 move control rods to correct position within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

D. Incorrect - LCO 3.1.6 Condition C states Be in MODE 3 if required actions not completed. This could be construed by the applicant since numerous rods are out of sequence. 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months is allowed to correct the condition.

LCO 3.1.6 Rod Pattern Control Technical Reference(s): rev. 224, LCO 3.1.3 Control Rod (Attach if not previously provided)

Operability rev. 271 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 33

Proposed References to be provided to applicants during examination: L3.1.6, L3.1.3 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 2 Comments: Facility operating limitations in the technical specifications and their bases.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 34

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 216000 A2.07 Importance Rating 3.5 Ability to (a) predict the impacts of the following on the NUCLEAR BOILER INSTRUMENTATION ; and (b) based on those predictions, use procedures to correct, control, or mitigate the consequences of those abnormal conditions or operations: Reference leg flashing Proposed Question: SRO Question 17 (92)

A normal plant Shutdown and Cooldown from full power operations is in progress.

The reactor is shutdown

RPV Pressure is 940 psig

RPV Level is 190 inches and stable

Cooldown has just started

The GEMAC Reference Leg Backfill System has been out of service for 3 weeks (1) What compensatory actions are directed to address these plant conditions?

AND (2) Which procedure is directed if NOTCHING is observed?

A. (1) When RPV pressure reaches 500 psig direct the operating crew NOT to use the Yarway instruments on 1C05 for level indication (2) OI 880, Non-Nuclear Instrumentation System B. (1) When RPV pressure reaches 500 psig direct the operating crew NOT to use the Yarway Instruments on 1C05 for level indication (2) IPOI 4, Shutdown C. (1) Direct enhanced RPV Level monitoring during the plant cooldown (2) OI 880, Non-Nuclear Instrumentation System D. (1) Direct enhanced RPV Level monitoring during the plant cooldown (2) IPOI 4, Shutdown Proposed Answer: C Explanation (Optional):

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 35

A. Incorrect - Yarways share instrument lines with the GEMAC instruments but these directions would be given for a rapid pressure decrease. OI 880 is the correct OI for the action required.

B. Incorrect - Yarways share instrument lines with the GEMAC instruments but these directions would be given for a rapid pressure decrease. IPOI 4 does direct action be performed IAW OI 880 for these conditions but does not have directions if Notching is observed.

C. Correct - Anytime the Reference Leg Backfill system is out of service for 7 days performance of OI-880, J-1 section 6.1 is required. Notching is expected when reducing RPV pressure during a shutdown. The Narrow range GEMAC level instruments are susceptible. OI-880 directs these actions for the given plant conditions.

D. Incorrect - IPOI 4 does direct action be performed IAW OI 880 for these conditions but does not have directions if Notching is observed. Enhanced level monitoring is the correct direction for these conditions.

OI-880 Non Nuclear Technical Reference(s): Instrumentation System section J- (Attach if not previously provided) 1, Rev.22 page 18 step 3 IPOI 4, Rev.

Proposed References to be provided to applicants during examination: N 4.18.03, Direct operator actions to control RPV level throughout the Learning Objective: (As available) cooldown.

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

New Question History: Last NRC Exam: 2009 NRC Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 5 Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 36

Examination Outline Cross-reference: Level RO SRO Tier # 2 Group # 2 K/A # 290001 G2.4.6 Importance Rating 4.7 Knowledge of EOP mitigation strategies.: Secondary Containment Proposed Question: SRO Question 18 (93)

With the plant operating at full power, a spent resin spill from the RWCU Phase Separator occurs in the reactor building. The following conditions are present:

RWCU Spent Resin Room ARM, RI-9173, is reading 3,500 mR/hr

RWCU Phase Separator Room ARM, RI-9177, is reading 850 mR/hr

North Refuel Floor ARM, RI-9163, is reading 210 mR/hr Which one of the following actions will be directed by the CRS?

A. Perform a fast power reduction and SCRAM the reactor B. Anticipate Emergency Depressurization C. Perform a plant shutdown per IPOI 3 D. Emergency Depressurize Proposed Answer: C Explanation (Optional):

A. A: Incorrect - would be true if reactor were discharging into the Secondary Containment (SC-3) . The conditions present represent a resin spill that would not be affected by lowering the energy state of the Reactor. The applicant may choose this condition due to exceeding Max Safe Operating Limits in ONE area per EOP 3, Secondary Containment Control/Rad Release Control.

B. Incorrect - would be true if reactor were discharging into the Secondary Containment (SC-3) and Max Safe Operating Limits were bing approached.

C. Correct Answer: It is required to perform a plant shutdown due to exceeding Max Safe Operating Limit in TWO areas. (SC-9)

D. Incorrect - would be true if reactor were discharging into the Secondary Containment.

(SC-7)

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 37

Bases- EOP 3 / EOP 4 Secondary Technical Reference(s): Containment / Rad Release (Attach if not previously provided)

Control, rev. 22 Proposed References to be provided to applicants during examination: EOP Table 6 6.68.02.01, Justify the use of IPOI 3,4, Learning Objective: or 5 to perform a plant shutdown while (As available) performing EOP 3.

Question Source: Bank #

Modified Bank # 52213 (Note changes or attach parent)

New PDA 17-1 EOP Question History: Last NRC Exam:

Comp Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4 Comments: Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 38

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G1 G2.1.5 Importance Rating 3.9 Ability to use procedures related to shift staffing, such as minimum crew complement, overtime limitations, etc.

Proposed Question: SRO Question 19 (94)

The plant is operating at full reactor power during a weekend night shift. The Shift Technical Advisor's (STA) wife just called to say she is headed to the hospital to have a baby. The STA asks to leave the site to accompany his wife prior to watch relief.

Is this permitted per ACP 1410.1, Operations Working Standard?

A. No, a qualified relief must first be on-site and relieve the STA.

B. Yes, the only action necessary is to obtain the Operations Director's approval to secure the STA for the remainder of the shift.

C. Yes, as long as an immediate callout is made for watch relief and the relief arrives within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The Operations Director must be notified.

D. Yes, as long as an immediate callout is made for watch relief and the relief arrives within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The Operations Director must be notified.

Proposed Answer: C Explanation (Optional):

A. Incorrect - It is allowed.

B. Incorrect - Ops Manager approval not necessary, just notification C. Correct - IAW T.S. 5.2.2: Shift crew composition may be less than the minimum requirement for a period of time not to exceed 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in order to accommodate unexpected absence of on-duty shift crew members provided immediate action is taken to restore the shift crew composition to within the minimum requirements. IAW ACP 1410.1, Operations Director must be notified.

D. Incorrect - An immediate call out is necessary, and the relief must be on site within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

ACP 1410.1, Rev. 104, Sect. 3.2, Technical Reference(s): Step (4) (a) & (b) LCO 5.2.2 (Attach if not previously provided)

Organization rev. 274 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 39

T.S. 5.2.2, Amendment 274 Proposed References to be provided to applicants during examination: N 1.11.03, ensure plant activities are Learning Objective: performed in compliance with the ACP (As available) 1400 Manual.

Question Source: Bank # 50878 Modified Bank # (Note changes or attach parent)

New PDA 13-1 SRO Question History: Last NRC Exam:

DAEC Cert Question Cognitive Level: Memory or Fundamental Knowledge Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 1 Comments: Conditions and limitations in the facility license.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 40

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G1 G2.1.36 Importance Rating 4.1 Knowledge of procedures and limitations involved in core alterations.

Proposed Question: SRO Question 20 (95)

Whose permission is required to commence Irradiated Fuel Handling?

A. Site Director B. Operations Shift Manager C. Control Room Supervisor D. Radiation Protection Manager Proposed Answer: B Explanation (Optional):

A. Incorrect - The Site Director permission is not required B. Correct - The OSM must approve commencement of fuel handling C. Incorrect - The CRS permission is not required D. Incorrect - The RPM persmission is not required RFP 403 rev 60 Attachment 1 step Technical Reference(s): (Attach if not previously provided) 21.

Proposed References to be provided to applicants during examination: N 1.04.01.03, describe the information found in the Fuel Moving Plan and Learning Objective: (As available) explain how minor changes are made to the Fuel Moving Plan, Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 41

Question Source: Bank #

Modified Bank # (Note changes or attach parent)

New X PDA 15-1 SRO Question History: Last NRC Exam:

Workup Comp A Question Cognitive Level: Memory or Fundamental Knowledge X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 6 Comments: 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.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 42

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2 G2.2.6 Importance Rating 3.6 Knowledge of the process for making changes to procedures.

Proposed Question: SRO Question 21 (96)

While performing a section of Operating Instruction (OI) 410, River Water Supply System, the reactor operator comes across this step:

(27) Place Traveling Screen Wash Pump 1P-112A[B] in service by placing handswitch HS-2906A[B] to the AUTO position on breaker 1B9106 [1B2106].

This step is unable to be performed. The Reactor Operator (RO) has determined performance of the step is NOT required.

Which one of the following meets the approval requirements to place an NA in the block for this step?

A. Two Senior Reactor Operators B. Only the Operations Shift Manager C. The STA and the RO performing the procedure D. The Control Room Supervisor and the RO performing the procedure Proposed Answer: A Explanation (Optional):

A. Correct: Two (2) SROs are required to review, approve, and initial an NA to a step in a work instruction or procedure that are for safety related equipment or equipment that supports technical specifications.

B. Incorrect: The OSM would need a second SRO C. Incorrect: Two SROs are required to perform this action D. Incorrect: The CRS would need a second SRO AD-AA-100-1006 Procedure Work Technical Reference(s): Instruction Use and Adherence, (Attach if not previously provided) page 28 section 4.8 step 5D, Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 43

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 3 Comments: Facility licensee procedures required to obtain authority for design and operating changes in the facility.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 44

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G2 2.2.44 Importance Rating 4.4 Ability to interpret control room indications to verify the status and l operation of a system, and understand how operator actions and directives affect plant and system conditions.

Proposed Question: SRO Question 22 (97)

The plant is operating at 100% power with the following conditions:

Instrument and Control Technicians determine that all Torus Water Level indications are indicating 6 inches higher than actual Torus Water Level.

Currently Torus Water Level indicates 10.35 Feet What is required?

A. Fill the Torus B. Scram the Reactor and enter EOP 1 C. Emergency Depressurize the RPV D. Increased monitoring of Torus Water Level indication Proposed Answer: A Explanation (Optional):

A. Correct - Torus water level actual is 9.85 Feet which is below the entry condition of 10.1 feet in EOP 2. LCO 3.6.2.2 applies. This condition must be corrected within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

B. Incorrect - This action is not required for the given conditions. Level is stable but low.

C. Incorrect - This action is not required for the given conditions. Level is stable but low.

D. Incorrect - This action in itself is not required. The required action is to fill the Torus to the appropriate level per Technical Specifications and EOP guidance.

EOP 2, TS LCO 3.6.2.2 Technical Reference(s): Suppression Pool Water Level rev. (Attach if not previously provided) 224 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 45

Proposed References to be provided to applicants during examination: L3.6.2.2 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 5 Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 46

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G3 G2.3.5 Importance Rating 2.9 Ability to use radiation monitoring systems, such as fixed radiation monitors and alarms, portable survey instruments, personnel monitoring equipment, etc.

Proposed Question: SRO Question 23 (98)

An accident has occurred which has caused entry into the RPV Flood EOP. All control rods are fully inserted.

Drywell area Rad monitors are now reading 450 and 500 R/Hr and slowly rising

Torus area Rad monitors are now reading 45 and 50 R/hr and slowly rising

The TSC is NOT manned at this time Given these plant conditions and EOP Table 5, which of the following actions is required at this time?

A. Continue to perform RPV/F actions. SAG entry is not required until multiple Core damage indications are seen B. Continue to perform RPV/F actions; since core flooding indications exist, enter the SAGs and then exit the EOPs C. Continue to perform RPV/F actions until RPV reactor water level indications are observed. Once the TSC is operational, exit the EOPs and enter the SAGs D. Enter the SAGs and transition from RPV/F to SAG 1, Primary Containment Flooding.

Continue to monitor the plant for degrading conditions and report them to the TSC once Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 47

manned and operational Proposed Answer: C Explanation (Optional):

A. Incorrect - RPV/F actions must continue however there is a SAG entry condition on primary containment radiation. Plausible because RPV/F action must continue and the candidate may misinterpret Table 5.

B. Incorrect - EOPs should not be exited until there is evidence of core flooding and the TSC is operational. With reactor pressure still lowering and SRV tailpipe temperatures stable there is NO evidence of core flooding, with the core NOT flooded and the TSC NOT operational the EOPs should NOT be exited at this time. Plausible because there is an SAG entry condition on primary containment radiation.

C. Correct - Based upon the containment Rad levels during RPV flooding fuel damage is occurring. This requires entry to SAGs. The transition to SAGs is performed by continuing the EOP actions until there is evidence that the core is flooded or reactor water level indication is available and the TSC is operational. Once the TSC is operational, The SAGs are entered at the appropriate point and directed by the TSC and then the EOPs are exited and EOP actions terminated.

D. Incorrect - EOPs should NOT be exited until the TSC is operational. The transition to SAGs is performed by continuing the EOP actions until there is evidence that the core is flooded.

RPV/F RPV Flooding , Rev. 15 Technical Reference(s): (Attach if not previously provided)

Table 5 Proposed References to be provided to applicants during examination:

6.85.06.01, discuss the process that is required to transition from EOPs to Learning Objective: (As available)

Severe Accident Management procedures Question Source: Bank # 52275 Modified Bank # (Note changes or attach parent)

New PDA 15-1 SRO Question History: Last NRC Exam:

Workup Comp A Question Cognitive Level: Memory or Fundamental Knowledge Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 48

Comprehension or Analysis X 10 CFR Part 55 Content: 55.41 55.43 4 Comments: Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 49

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G4 G2.4.37 Importance Rating 4.1 Knowledge of the lines of authority during implementation of the emergency plan.

Proposed Question: SRO Question 24 (99)

An event has resulted in an unisolable leak and lowering level in the fuel pool.

The Shift manager declared an ALERT due to abnormal rising radiation levels on the refuel floor 60 minutes ago.

All onsite and offsite personnel have responded to their designated emergency response organizations.

The Technical Support Center is staffed but not yet operational.

The EOF is staffed but not yet operational.

It has been determined a task on the refuel floor is required to be performed by two operators.

Which one of the following personnel are required to approve access to the refuel floor?

A. Shift Manager B. Site Director C. Site Radiation Protection Coordinator D. Emergency Response and Recovery Director Proposed Answer: A Explanation (Optional):

A. Correct - EPIP 2.5 Control Room Emergency Response Operation, Revision 21 page 11: If suspected abnormal radiation conditions exist the Emergency Coordinator must approve the dispatch of all personnel into the plant.The SM is the EC until the TSC is fully staffed and has command and control B. Incorrect - The Site Director is not the listed approval required C. The SRPC, while this person does assess radiation conditions on and off site, the decision is made be the EC D. The ERRD is not in command of the event at this point EPIP 2.5 Control Room Technical Reference(s): Emergency Response Operation, (Attach if not previously provided)

Revision 21 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 50

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 X Comprehension or Analysis 10 CFR Part 55 Content: 55.41 55.43 4 Comments: Radiation hazards that may arise during normal and abnormal situations, including maintenance activities and various contamination conditions.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 51

Examination Outline Cross-reference: Level RO SRO Tier #

Group #

K/A # G4 G2.4.38 Importance Rating 4.4 Ability to take actions called for in the facility emergency plan, including supporting or acting as emergency coordinator if required.

Proposed Question: SRO Question 25 (100)

An event has occurred at the DAEC and the following conditions are now present:

A Steam leak in the Steam Tunnel has caused temperatures to rise to 250°F

Steam Tunnel Temperatures continue to rise

The A Main Steam Line Isolation Valves failed to close automatically and will not close from the control room Which one of the following EAL declarations should be made?

A. FU1 B. FA1 C. FS1 D. FG1 Proposed Answer: C Explanation (Optional):

A. Incorrect: Plausible in that the candidate may determine that only the Primary Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 52

Containment barrier has failed.

B. Incorrect: Plausible in that the candidate may determine that only RCS barrier was lost or potentially lost requiring declaration of FS1. The Fuel Clad Barrier remains intact.

C. Correct: EPIP EAL Hot Matrix Failure and loss of the primary containment barrier and RCS Barrier resulting D. Incorrect: Plausible in that the candidate may determine all 3 fission product barriers are lost.

Technical Reference(s): EPIP 1.1 Determination of (Attach if not previously provided)

Emergency Action Levels, Rev. 29 EPIP Form EAL-01 Proposed References to be provided to applicants during examination:

3.01.01.01, Explain the Learning Objective: (As available)

Responsibilities and Instructions contained in EPIP 1.1 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 Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 53

55.43 5 Comments: Assessment of facility conditions and selection of appropriate procedures during normal, abnormal, and emergency situations.

Rev. 0 60006_PDA 17-1 NRC Exam SRO_xm_60Day.docx Page 54

ML17214A050

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